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Sample records for high pressure coolant injection

  1. High Pressure Coolant Injection (HPCI) system risk-based inspection guide: Pilgrim Nuclear Power Station

    International Nuclear Information System (INIS)

    Shier, W.; Gunther, W.

    1992-10-01

    A review of the operating experience for the High Pressure Coolant Injection (HPCI) system at the Pilgrim Nuclear Power Station is described in this report. The information for this review was obtained from Pilgrim Licensee Event Reports (LERs) that were generated between 1980 and 1989. These LERs have been categorized into 23 failure modes that have been prioritized based on probabilistic risk assessment considerations. In addition, the results of the Pilgrim operating experience review have been compared with the results of of a similar, industry wide operating experience review. this comparison provides an indication of areas in the Pilgrim HPCI system that should be given increased attention in the prioritization of inspection resources

  2. High Pressure Coolant Injection system risk-based inspection guide for Hatch Nuclear Power Station

    International Nuclear Information System (INIS)

    DiBiasio, A.M.

    1993-05-01

    A review of the operating experience for the High Pressure Coolant Injection (HPCI) system at the Hatch Nuclear Power Station, Units 1 and 2, is described in this report. The information for this review was obtained from Hatch Licensee Event Reports (LERs) that were generated between 1980 and 1992. These LERs have been categorized into 23 failure modes that have been prioritized based on probabilistic risk assessment considerations. In addition, the results of the Hatch operating experience review have been compared with the results of a similar, industry wide operating, experience review. This comparison provides an indication of areas in the Hatch HPCI system that should be given increased attention in the prioritization of inspection resources

  3. Browns Ferry Nuclear Plant: variation in test intervals for high-pressure coolant injection (HPCI) system

    International Nuclear Information System (INIS)

    Christie, R.F.; Stetkar, J.W.

    1985-01-01

    The change in availability of the high-pressure coolant injection system (HPCIS) due to a change in pump and valve test interval from monthly to quarterly was analyzed. This analysis started by using the HPCIS base line evaluation produced as part of the Browns Ferry Nuclear Plant (BFN) Probabilistic Risk Assessment (PRA). The base line evaluation showed that the dominant contributors to the unavailability of the HPCI system are hardware failures and the resultant downtime for unscheduled maintenance. The effect of changing the pump and valve test interval from monthly to quarterly was analyzed by considering the system unavailability due to hardware failures, the unavailability due to testing, and the unavailability due to human errors that potentially could occur during testing. The magnitude of the changes in unavailability affected by the change in test interval are discussed. The analysis showed a small increase in the availability of the HPCIS to respond to loss of coolant accidents (LOCAs) and a small decrease in the availability of the HPCIS to respond to transients which require HPCIS actuation. In summary, the increase in test interval from monthly to quarterly does not significantly impact the overall HPCIS availability

  4. High Pressure Coolant Injection (HPCI) System Risk-Based Inspection Guide for Browns Ferry Nuclear Power Station

    International Nuclear Information System (INIS)

    Wong, S.; DiBiasio, A.; Gunther, W.

    1993-09-01

    The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A System Risk-Based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Browns Ferry Nuclear Power Plant, Units 1, 2 and 3. The role of. the HPCI system in mitigating accidents is discussed in this S-RIG, along with insights on identified risk-based failure modes which could prevent proper operation of the system. The S-RIG provides a review of industry-wide operating experience, including plant-specific illustrative examples to augment the PRA and operational considerations in identifying a catalogue of basic PRA failure modes for the HPCI system. It is designed to be used as a reference for routine inspections, self-initiated safety system functional inspections (SSFIs), and the evaluation of risk significance of component failures at the nuclear power plant

  5. High Pressure Coolant Injection (HPCI) System Risk-Based Inspection Guide for Browns Ferry Nuclear Power Station

    Energy Technology Data Exchange (ETDEWEB)

    Wong, S.; DiBiasio, A.; Gunther, W. [Brookhaven National Lab., Upton, NY (United States)

    1993-09-01

    The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A System Risk-Based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Browns Ferry Nuclear Power Plant, Units 1, 2 and 3. The role of. the HPCI system in mitigating accidents is discussed in this S-RIG, along with insights on identified risk-based failure modes which could prevent proper operation of the system. The S-RIG provides a review of industry-wide operating experience, including plant-specific illustrative examples to augment the PRA and operational considerations in identifying a catalogue of basic PRA failure modes for the HPCI system. It is designed to be used as a reference for routine inspections, self-initiated safety system functional inspections (SSFIs), and the evaluation of risk significance of component failures at the nuclear power plant.

  6. Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

    2003-01-01

    To support the development of a Probabilistic Safety Assessment (PSA) model usable in Riskinformed Applications (RIA) for Korea Standard Nuclear power Plants (KSNP), we have performed a thermal hydraulic analysis of Aggressive Secondary Cooldown (ASC) in a 2-inch Small Break Loss Of Coolant Accident (SBLOCA) with a total loss of High Pressure Safety Injection (HPSI). The present study focuses on the estimation of the success criteria of ASC, and the enhanced understanding of the detailed thermal hydraulic behavior and phenomena. The results have shown that the Reactor Coolant System (RCS) pressure can be reduced to the Low Pressure Safety Injection (LPSI) operation conditions without core damage. It was also shown that more relaxed success criteria compared to those in the previous PSA models of KSNP could be used in the new PSA model. However, it was found that the results could be affected by various parameters related with ASC operation, i.e., reference temperature for the calculation of the cooldown rate and its control method

  7. Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 3. Behaviour of high pressure coolant injection system (HPCI) based on thermodynamic model

    International Nuclear Information System (INIS)

    Maruyama, Shigenao

    2014-01-01

    In order to clarify the process of Accident of Fukushima Nuclear Plants, an accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 3 is analyzed from the data open to the public. Phase equilibrium process model was introduced in which the vapor and water are at saturation point in the vessels. The present accident scenario assumes that the high pressure coolant injection system (HPCI) did not worked properly, but the steam in the reactor pressure vessel (RPV) leaked through the turbine of HPCI to the suppression chamber since 12/3/2011 12:35. It is assumed that the Tsunami flooded the torus room where the suppression chamber was placed. Proposed accident scenario agrees with the data of the plant parameters obtained just after the accident. It is estimated that the water injection by HPIC was stopped since around at 13/3 19:00 and the water level in RPV decreased since then. It is estimated that the RPV broke at 14/3 8:55 and water could injected from fire engines due to the depression due to the rupture of RPV. There was little water left in RPV at the time of the rupture. If the present scenario is correct, the behavior that operators in the plant stopped HPCI at 13/3 2:42 did not affect seriously on the RPV rupture. If HPCI was working properly until the operators stopped it, the plant parameters obtained in the accident cannot be explained. (author)

  8. Thermal hydraulic analysis of aggressive secondary cooldown in small break loss of coolant accident with total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, S. J.; Im, H. K.; Yang, J. U.

    2003-01-01

    Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). To use RIA, the present study focuses on the detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study is to evaluate the success criteria of Aggressive Secondary Cooldown (ASC) in Small Break Loss Of Coolant Accident (SBLOCA) with total loss of High Pressure Safety Injection (HPSI) and to enhance the understanding of related thermal hydraulic behavior and phenomena. The accident scenario was 2 inch coldleg break LOCA without HPSI, with 1/2 Low Pressure Safety Injection (LPSI), and performing ASC limited by 55.6 .deg. C /hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip, which successively reaches the LPSI condition for about 1.5hr after starting ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria 1204.4 .deg. C (2200 .deg. F). In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that operator should maintain the adequate ASC operation. However, it is necessary to evaluate uncertainties arisen from the related parameters of the ASC operation

  9. Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

    2003-03-01

    Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). The present study focuses on detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model using RIA for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study in this year is to evaluate the success cri-teria of Aggressive Secondary Cooldown (ASC) in a Small Size Loss Of Coolant Accident (SBLOCA) without HPSI and to enhance the understanding of related thermal hydraulic behavior and phenomena. An effort was made to evaluate the system success criteria and a mission time for the recovery action by an operator to prevent the core damage for that accident scenario. The accident scenario for KSNP was a 2 inch coldleg break LOCA with a total loss of High Pressure Safety Injection (HPSI) and 1/2 Low Pressure Safety Injection (LPSI) available and perform-ing ASC limited by 55.6 .deg. C/hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip. It successively reached the LPSI condition for about 1.5hr after starting the ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria of 1204.4 .deg. C (2200 .deg. F). Sensitivity studies were performed for (1) cool-ant average temperature parameters, (2) ASC operation control method, (3) operation start time, (4) 1 inch break size. The present analysis identified thermal hydraulic phenomena and parameters affecting on the behavior, which consist of coolant break flow and inventory, parameters governing secondary heat removal, ASC operation control method, and its reference temperature parameters. In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that an operator should maintain the ade-quate ASC operation. However, it is necessary to evaluate the uncertainties arisen from the

  10. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  11. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  13. Aging study of boiling water reactor high pressure injection systems

    International Nuclear Information System (INIS)

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

  14. High pressure injection of dimethyl ether

    Energy Technology Data Exchange (ETDEWEB)

    Glensvig, M.; Sorenson, S.C.; Abata, D.L.

    1997-08-01

    The purpose of this investigation was to achieve a better understanding of the fundamental spray behavior of DME (Dimenthyl Ether) using a standard diesel pump with pintle and hole nozzles. Fundamental spray behavior was characterized by determining fuel spray penetration and angle, atomization and evaporation. The influences of opening pressure, nozzle geometry and ambient pressure above and below the critical pressure of the fuel on the spray behavior were investigated. The influence of opening pressures on the spray characteristics for the hole nozzle was investigated. The results showed that for opening pressures of 120 bar and 180 bar the spray has a similar appearance. For the higher opening pressure (200 bar and 240 bar), the initial spray breaks up very rapidly giving a high initial spray angle. The opening pressure had little influence on spray penetration. The spray angle later in the injection increased as the opening pressure was decreased. Above the critical pressure, the spray from the hole nozzle had a more irregular shape. Penetration decreased and the spray angle increased above the critical pressure. Three pintle nozzles with different geometries and opening pressures were tested. The appearance of the three sprays were very similar. The sprays seemed to be more sharply pointed as the nozzle hole angle decreased. The nozzle with the 4 deg. hole nozzle angle and an opening pressure of 280 bar had the highest penetration and highest initial spray angle. The pintle nozzle with the 12 deg. hole nozzle angle and opening pressure of approx. 450 bar was tested above the critical ambient pressure. Penetration was very similar for injection above and below the critical ambient pressure, while the spray angle decreased for the spray above the critical ambient pressure. (au)

  15. High-pressure coolant effect on the surface integrity of machining titanium alloy Ti-6Al-4V: a review

    Science.gov (United States)

    Liu, Wentao; Liu, Zhanqiang

    2018-03-01

    Machinability improvement of titanium alloy Ti-6Al-4V is a challenging work in academic and industrial applications owing to its low thermal conductivity, low elasticity modulus and high chemical affinity at high temperatures. Surface integrity of titanium alloys Ti-6Al-4V is prominent in estimating the quality of machined components. The surface topography (surface defects and surface roughness) and the residual stress induced by machining Ti-6Al-4V occupy pivotal roles for the sustainability of Ti-6Al-4V components. High-pressure coolant (HPC) is a potential choice in meeting the requirements for the manufacture and application of Ti-6Al-4V. This paper reviews the progress towards the improvements of Ti-6Al4V surface integrity under HPC. Various researches of surface integrity characteristics have been reported. In particularly, surface roughness, surface defects, residual stress as well as work hardening are investigated in order to evaluate the machined surface qualities. Several coolant parameters (including coolant type, coolant pressure and the injection position) deserve investigating to provide the guidance for a satisfied machined surface. The review also provides a clear roadmap for applications of HPC in machining Ti-6Al4V. Experimental studies and analysis are reviewed to better understand the surface integrity under HPC machining process. A distinct discussion has been presented regarding the limitations and highlights of the prospective for machining Ti-6Al4V under HPC.

  16. High pressure injection injuries: an overview.

    Science.gov (United States)

    Fialkov, J A; Freiberg, A

    1991-01-01

    Injuries resulting from the use of high pressure injectors and spray guns are relatively rare; however, the potential tissue damage caused by the injury as well as the extent of the injury itself may go unrecognized by the primary physician. The purpose of this paper is to inform the emergency physician of the nature and standard management of this type of injury. A basic understanding of the pathophysiology of the high pressure injection injury (HPII) is essential in avoiding the mistakes in management that have been reported in the literature. The emergency management of the HPII includes: evaluation and immobilization, tetanus and antimicrobial prophylaxis, supportive and resuscitative measures, analgesia, and minimizing the time to definitive surgical treatment.

  17. Thermal shock studies associated with injection of emergency core coolant in pressurized water reactors

    International Nuclear Information System (INIS)

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

    1977-01-01

    Studies to determine the accuracy of calculational techniques for predicting crack initiation and arrest in PWR vessels due to thermal shock from ECC injection are described. The reference calculational model is reviewed, the experimental program and facilities are described, and some thermal shock experiments and results are discussed

  18. Impact of high-pressure coolant supply on chip formation in milling

    Science.gov (United States)

    Klocke, F.; Döbbeler, B.; Lakner, T.

    2017-10-01

    Machining of titanium alloys is considered as difficult, because of their high temperature strength, low thermal conductivity and low E-modulus, which contributes to high mechanical loads and high temperatures in the contact zone between tool and workpiece. The generated heat in the cutting zone can be dissipated only in a low extent. When cutting steel materials, up to 75% of the process heat is transported away by the chips, contrary to only 25% when machining titanium alloys. As a result, the cutting tool heats up, which leads to high tool wear. Therefore, machining of titanium alloys is only possible with relatively low cutting speeds. This leads to low levels of productivity for milling processes with titanium alloys. One way to increase productivity is to use more cutting edges in tools with the same diameter. However, the limiting factor of adding more cutting edges to a milling tool is the minimum size of the chip spaces, which are sufficient for a stable chip evacuation. This paper presents experimental results on the chip formation and chip size influenced by high-pressure coolant supply, which can lead to smaller chips and to smaller sizes of the chip spaces, respectively. Both influences, the pressure of the supplied coolant and the volumetric flow rate were individually examined. Alpha-beta annealed titanium TiAl6V4 was examined in relation to the reference material quenched and tempered steel 42CrMo4+QT (AISI 4140+QT). The work shows that with proper chip control due to high-pressure coolant supply in milling, the number of cutting edges on the same diameter tool can be increased, which leads to improved productivity.

  19. Analysis of thermo-hydraulic behavior of coolant during discharge of pressurized high-temperature water

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Sobajima, Makoto; Sasaki, Shinobu; Onishi, Nobuaki; Shiba, Masayoshi

    1978-01-01

    The present report describes results of the analysis of the LOFT semiscale experiment No. 1011 using remodeled RELAP-3 code, performed at the Idaho National Engineering Laboratory to simulate a postulated loss-of-coolant accident in a pressurized water reactor. It was clarified through the analysis that coolant behavior during blowdown was influenced variously by the system components in the primary loop, comparing with coolant discharge from a pressure vessel. Good agreement was obtained between experimental and analytical results when phase separation was assumed in upper plenum and downcomer, since experimental data indicated existence of liquid level in those parts. It was also found that the use of the Wilson's equation to calculate bubble rise velocity and the use of discharge coefficient as the function of fluid quality at break location to calculate discharge flow rate resulted in good agreement with experimental data. (auth.)

  20. High converter pressurized water reactor with heavy water as a coolant

    International Nuclear Information System (INIS)

    Ronen, Y.; Reyev, D.

    1983-01-01

    There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

  1. Pneumomediastinum following high pressure air injection to the hand.

    LENUS (Irish Health Repository)

    Kennedy, J

    2012-02-01

    We present the case of a patient who developed pneumomediastinum after high pressure air injection to the hand. To our knowledge this is the first reported case of pneumomediastinum where the gas injection site was the thenar eminence. Fortunately the patient recovered with conservative management.

  2. Pneumomediastinum following high pressure air injection to the hand.

    LENUS (Irish Health Repository)

    Kennedy, J

    2010-04-01

    We present the case of a patient who developed pneumomediastinum after high pressure air injection to the hand. To our knowledge this is the first reported case of pneumomediastinum where the gas injection site was the thenar eminence. Fortunately the patient recovered with conservative management.

  3. Computational Fluid Dynamics Analysis of High Injection Pressure Blended Biodiesel

    Science.gov (United States)

    Khalid, Amir; Jaat, Norrizam; Faisal Hushim, Mohd; Manshoor, Bukhari; Zaman, Izzuddin; Sapit, Azwan; Razali, Azahari

    2017-08-01

    Biodiesel have great potential for substitution with petrol fuel for the purpose of achieving clean energy production and emission reduction. Among the methods that can control the combustion properties, controlling of the fuel injection conditions is one of the successful methods. The purpose of this study is to investigate the effect of high injection pressure of biodiesel blends on spray characteristics using Computational Fluid Dynamics (CFD). Injection pressure was observed at 220 MPa, 250 MPa and 280 MPa. The ambient temperature was kept held at 1050 K and ambient pressure 8 MPa in order to simulate the effect of boost pressure or turbo charger during combustion process. Computational Fluid Dynamics were used to investigate the spray characteristics of biodiesel blends such as spray penetration length, spray angle and mixture formation of fuel-air mixing. The results shows that increases of injection pressure, wider spray angle is produced by biodiesel blends and diesel fuel. The injection pressure strongly affects the mixture formation, characteristics of fuel spray, longer spray penetration length thus promotes the fuel and air mixing.

  4. High pressure common rail injection system modeling and control.

    Science.gov (United States)

    Wang, H P; Zheng, D; Tian, Y

    2016-07-01

    In this paper modeling and common-rail pressure control of high pressure common rail injection system (HPCRIS) is presented. The proposed mathematical model of high pressure common rail injection system which contains three sub-systems: high pressure pump sub-model, common rail sub-model and injector sub-model is a relative complicated nonlinear system. The mathematical model is validated by the software Matlab and a virtual detailed simulation environment. For the considered HPCRIS, an effective model free controller which is called Extended State Observer - based intelligent Proportional Integral (ESO-based iPI) controller is designed. And this proposed method is composed mainly of the referred ESO observer, and a time delay estimation based iPI controller. Finally, to demonstrate the performances of the proposed controller, the proposed ESO-based iPI controller is compared with a conventional PID controller and ADRC. Copyright © 2016 ISA. Published by Elsevier Ltd. All rights reserved.

  5. Effect of cavitation in high-pressure direct injection

    Science.gov (United States)

    Aboulhasanzadeh, Bahman; Johnsen, Eric

    2015-11-01

    As we move toward higher pressures for Gasoline Direct Injection and Diesel Direct Injection, cavitation has become an important issue. To better understand the effect of cavitation on the nozzle flow and primary atomization, we use a high-order accurate Discontinuous Galerkin approach using multi-GPU parallelism to simulate the compressible flow inside and outside the nozzle. Phase change is included using the six-equations model. We investigate the effect of nozzle geometry on cavitation inside the injector and on primary atomization outside the nozzle.

  6. Light extinction method on high-pressure diesel injection

    Science.gov (United States)

    Su, Tzay-Fa; El-Beshbeeshy, Mahmound S.; Corradini, Michael L.; Farrell, Patrick V.

    1995-09-01

    A two dimensional optical diagnostic technique based on light extinction was improved and demonstrated in an investigation of diesel spray characteristics at high injection pressures. Traditional light extinction methods require the spray image to be perpendicular to the light path. In the improved light extinction scheme, a tilted spray image which has an angle with the light path is still capable of being processed. This technique utilizes high speed photography and digital image analysis to obtain qualitative and quantitative information of the spray characteristics. The injection system used was an electronically controlled common rail unit injector system with injection pressures up to 100 MPa. The nozzle of the injector was a mini-sac type with six holes on the nozzle tip. Two different injection angle nozzles, 125 degree(s) and 140 degree(s), producing an in-plane tilted spray and an out of plane tilted spray were investigated. The experiments were conducted on a constant volume spray chamber with the injector mounted tilted at an angle of 62.5 degree(s)$. Only one spray plume was viewed, and other sprays were free to inject to the chamber. The spray chamber was pressurized with argon and air under room temperature to match the combustion chamber density at the start of the injection. The experimental results show that the difference in the spray tip penetration length, spray angle, and overall average Sauter mean diameter is small between the in- plane tilted spray and the out of plane tilted spray. The results also show that in-plane tilted spray has a slightly larger axial cross- section Sauter mean diameter than the out of plane tilted spray.

  7. High-Pressure Injection Injuries to the Hand

    Directory of Open Access Journals (Sweden)

    Davod Jafari

    2016-07-01

    Full Text Available Background High-pressure injections into the hand, burden devastating and permanent functional impairments. Many materials including paint, paint thinner, gasoline, oil and grease are reported as the causative agents. These injuries need multiple procedures and reconstructions most of the time and 40% of the injuries may end with amputation of the injured part. Objectives The aim of this study was to report the treatment outcomes and methods of treatments of patients with high-pressure injection injuries of the hand. Methods We retrospectively reviewed the medical records, imaging files and demographic data of patients, who were treated at our center due to the high-pressure injuries to their hands. We recorded the kind of the injected materials, time to the first treatment procedure, times of operation, and methods of their treatments. The outcomes of the injuries as well as the deficiency of the digital joints motion were also reported. Results Nine cases with high-pressure injury of the hand were enrolled in this study. All patients were male with mean age of 26.88 ± 7.52. Mean follow-up time was 28.55 ± 12.49 months. The dominant hand was the right side in seven patients and left in two patients. Injury was in the left hand of seven patients and right hand of two patients. Index finger was the most common involved part (five cases followed by the thumb (two cases. Injected material was grease in seven cases, water-base paint and water, each in one case.Mean time delay to the first treatment procedure was 29.16 ± 25.66 hours for seven patients. This was exceptionally long for two patients (seven days and 24 months. Type of treatment was debridement and skin graft for three cases, debridement and cross finger flap for two cases, debridement for two cases and nerve graft for one case. Amputation of the necrotic digit was performed for one case. Mean hospitalization time was 8.33 ± 3.64 days for all patients.Mean total active range of motion

  8. System Study: High-Pressure Safety Injection 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-01

    This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPSI results.

  9. Primary break with total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Cordelle, F.; Champ, M.; Pochard, R.

    1988-10-01

    The probabilitic safety assessment of a 900 MW plant has displayed the potential importance, with regard to the risk, of intermediate primary breaks with failure of the high pressure safety injection system. The probability of such sequence is about 10 -6 /plant X year. Therefore, it is necessary to establish: - if this sequence can lead to core melt down, - if clad ruptures can occur. This event must be taken into account to determine the repair time of contaminated systems. For these studies, a three inch equivalent diameter break is considerd, as this is the most sensitive in its category with regard to these phenomena. In addition to the above objectives, the purpose of these studies is to evaluate the sensitivity of the results to the following parameters: - the time limit at which the operator starts cooling down the plant via the steam generators. Two calculations have been made with the RELAP code (1 and 2) and two with the CATHARE code (3 and 4) - the pump trip time. Four calculations have been made with the CATHARE code (5, 6, 7 and 8). In the case of failure of only one high pressure safety injection file, 6 calculations have been made with the CATHARE code, concerning the influence of pump trip time (9, 10, 11, 12, 13 and 14)

  10. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  11. [Results of treatment for high-pressure injection hand injuries].

    Science.gov (United States)

    Zyluk, A; Walaszek, I

    2000-01-01

    High-pressure injection injuries of the hand have a reputation for being dangerous for individual fingers and even for whole hand. Usually appearing innocuous at presentation because of small puncture entry wound, these injuries result in severe damage of most internal structures in finger and hand due to extensive penetration of injected substance. This paper reviews the outcome of the treatment of such injuries in 10 patients: 9 sustained injection of toxic paint, and one lead shot. All the patients were operated on: eight a few hours after injury and two with 3 days delay. The surgical technique included wide exposure from site of injection up to the farthest place in which foreign substance was seen. Thorough debridment of injected material and contaminated tissue was performed with careful preservation of neurovascular structures and tendons. Wounds were not closed, but managed by open technique. In all patients wounds healed well: in 3 by secondary intention, in 6 by delayed closure and 2 were covered by skin grafts. No amputation was performed. Final results were assessed form 1.5 to 3.5 years after initial injury (mean at 2.5 years). Two patients complained of moderate pain related to the weather, five of cold intolerance and two of impaired sensation on fingertips. Active range of motion of affected fingers was in whole group from 90% to 104% (mean 97%) of the range of motion of unaffected fingers from the other side. Range of motion of the wrist (2 patients) was 76% and 117% of range of motion of the other side. Pinch grip strength was from 81% to 116% (mean 99%), and global grip strength from 77% to 119% (mean 97%) of the other side. All patients went back to their previous jobs and periods of sick leave were from 2 weeks to 6 months (mean 3 mo). Excellent results achieved in this study--full functional recovery in 9 of 10 patients confirm the effectiveness of aggressive treatment by open wound technique of such injuries.

  12. Long-term follow-up of high-pressure injection injuries to the hand

    NARCIS (Netherlands)

    Wieder, Anat; Lapid, Oren; Plakht, Ygal; Sagi, Amiram

    2006-01-01

    High-pressure injection injury is an injury caused by accidental injection of substances by industrial equipment. This injury may have devastating sequelae. The goal of this study was to assess the long-term outcome of high-pressure injection injury to the hand. In this historical prospective study,

  13. Reflooding phase after loss of coolant of an advanced pressurized water reactor with high conversion ratio

    International Nuclear Information System (INIS)

    Schumann, S.

    1984-01-01

    The emergency core cooling behaviour of an advanced pressurized water reactor (APWR) during the reflooding phase of the LOCA with double-ended break is analysed and compared to a common pressurized water reactor (PWR). The code FLUT-BS, its models and correlations are explained in detail and have been verified by numerous PWR-reflood experiments with large parameter range. The influence of core-design on ECC-behaviour as well as the influences of initial and boundary values are examined. The results show the essential differences of ECC-behaviour between PWR and APWR. (orig.) [de

  14. Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

    International Nuclear Information System (INIS)

    Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

    2011-01-01

    Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

  15. Characteristics of pressure wave in common rail fuel injection system of high-speed direct injection diesel engines

    Directory of Open Access Journals (Sweden)

    Mohammad Reza Herfatmanesh

    2016-05-01

    Full Text Available The latest generation of high-pressure common rail equipment now provides diesel engines possibility to apply as many as eight separate injection pulses within the engine cycle for reducing emissions and for smoothing combustion. With these complicated injection arrangements, optimizations of operating parameters for various driving conditions are considerably difficult, particularly when integrating fuel injection parameters with other operating parameters such as exhaust gas recirculation rate and boost pressure together for evaluating calibration results. Understanding the detailed effects of fuel injection parameters upon combustion characteristics and emission formation is therefore particularly critical. In this article, the results and discussion of experimental investigations on a high-speed direct injection light-duty diesel engine test bed are presented for evaluating and analyzing the effects of main adjustable parameters of the fuel injection system on all regulated emission gases and torque performance. Main injection timing, rail pressure, pilot amount, and particularly pilot timing have been examined. The results show that optimization of each of those adjustable parameters is beneficial for emission reduction and torque improvement under different operating conditions. By exploring the variation in the interval between the pilot injection and the main injection, it is found that the pressure wave in the common rail has a significant influence on the subsequent injection. This suggests that special attentions must be paid for adjusting pilot timing or any injection interval when multi-injection is used. With analyzing the fuel amount oscillation of the subsequent injections to pilot separation, it demonstrates that the frequency of regular oscillations of the actual fuel amount or the injection pulse width with the variation in pilot separation is always the same for a specified fuel injection system, regardless of engine speed

  16. Evaluation of a coolant injection into the in-vessel with a RCS depressurization by using SCDAP/RELAP5

    International Nuclear Information System (INIS)

    Rae-Joon, Park; Sang-Baik, Kim; Hee-Dong, Kim

    2007-01-01

    As part of the evaluations of a severe accident management strategy, a coolant injection in the vessel with a reactor coolant system (RCS) depressurization has been evaluated by using the SCDAP/RELAP5 computer code. Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feed water (LOFW) accident have been analyzed in optimized power reactor OPR-1000. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 seconds with a RCS depressurization by using one condenser dump valve at 6 minutes after an entrance of the severe accident management guidance prevents a reactor vessel failure for the small break LOCA without SI. In this case, only train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent a reactor vessel failure. Only one train operation of the HPSI at 20,208 seconds with a RCS depressurization by using two safety depressurization system valves at 40 minutes after an initial opening of the safety relief valve prevents a reactor vessel failure for the total LOFW. (authors)

  17. High cyclic fatigue of PWR primary piping generated by the pressure pulsations in coolant

    International Nuclear Information System (INIS)

    Zd'arek, J.; Pecinka, L.; Zeman, V.

    1999-01-01

    The protection of nuclear piping Class 1, 2 and 3 against fatigue failure is according to standard western practise and is based on - determining the cumulative usage factor (CUF) using equation (11) of ASME Code, Section III, Article NB 3653 for Class 1 piping; - Markl experiments and equation (10) of ASME Code, Section III, Article NC/ND 3653 for Class 2/3 piping. These evaluations cover only low cyclic loading and the possible influence of high cyclic loading as for example vibratory stresses generated by the main circulating pumps are not taken into account. This problem is fully covered in the Czech and Russian codes. The goal of this paper is 1. to clarify the basic principles; 2. to discuss in detail the methodology for the calculation of high frequency vibratory stresses; and 3. to demonstrate with a numerical example, the degree of influence of the CUF. (orig.)

  18. High performance experiments on high pressure supersonic molecular beam injection in the HL-1M tokamak

    International Nuclear Information System (INIS)

    Yao Lianghua; Dong Jiafu; Zhou Yan; Feng Beibing; Cao Jianyong; Li Wei; Feng Zhen; Zhang Jiquan; Hong Wenyu; Cui Zhengying; Wang Enyao; Liu Yong

    2004-01-01

    Supersonic molecular beam injection (SMBI) was first proposed and demonstrated on the HL-1 tokamak and was successfully developed and used on HL-1M. Recently, new results of SMBI experiments were obtained by increasing the gas pressure from 0.5 to over 1.0 MPa. A stair-shaped density increment was obtained with high-pressure multi-pulse SMBI that was similar to the density evolution behaviour during multi-pellet injection. This demonstrated the effectiveness of SMBI as a promising fuelling tool for steady-state operation. The penetration depth and injection speed of the high-pressure SMBI were roughly measured from the contour plot of the Hα emission intensity. It was shown that injected particles could penetrate into the core region of the plasma. The penetration speed of high-pressure SMBI particles in the plasma was estimated to be about 1200 m s -1 . In addition, clusters within the beam may play an important role in the deeper injection. (author)

  19. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  20. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  1. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  2. Diesel Combustion and Emission Using High Boost and High Injection Pressure in a Single Cylinder Engine

    Science.gov (United States)

    Aoyagi, Yuzo; Kunishima, Eiji; Asaumi, Yasuo; Aihara, Yoshiaki; Odaka, Matsuo; Goto, Yuichi

    Heavy-duty diesel engines have adopted numerous technologies for clean emissions and low fuel consumption. Some are direct fuel injection combined with high injection pressure and adequate in-cylinder air motion, turbo-intercooler systems, and strong steel pistons. Using these technologies, diesel engines have achieved an extremely low CO2 emission as a prime mover. However, heavy-duty diesel engines with even lower NOx and PM emission levels are anticipated. This study achieved high-boost and lean diesel combustion using a single cylinder engine that provides good engine performance and clean exhaust emission. The experiment was done under conditions of intake air quantity up to five times that of a naturally aspirated (NA) engine and 200MPa injection pressure. The adopted pressure booster is an external supercharger that can control intake air temperature. In this engine, the maximum cylinder pressure was increased and new technologies were adopted, including a monotherm piston for endurance of Pmax =30MPa. Moreover, every engine part is newly designed. As the boost pressure increases, the rate of heat release resembles the injection rate and becomes sharper. The combustion and brake thermal efficiency are improved. This high boost and lean diesel combustion creates little smoke; ISCO and ISTHC without the ISNOx increase. It also yields good thermal efficiency.

  3. Picosecond ballistic imaging of diesel injection in high-temperature and high-pressure air

    Science.gov (United States)

    Duran, Sean P.; Porter, Jason M.; Parker, Terence E.

    2015-04-01

    The first successful demonstration of picosecond ballistic imaging using a 15-ps-pulse-duration laser in diesel sprays at temperature and pressure is reported. This technique uses an optical Kerr effect shutter constructed from a CS2 liquid cell and a 15-ps pulse at 532 nm. The optical shutter can be adjusted to produce effective imaging pulses between 7 and 16 ps. This technique is used to image the near-orifice region (first 3 mm) of diesel sprays from a high-pressure single-hole fuel injector. Ballistic imaging of dodecane and methyl oleate sprays injected into ambient air and diesel injection at preignition engine-like conditions are reported. Dodecane was injected into air heated to 600 °C and pressurized to 20 atm. The resulting images of the near-orifice region at these conditions reveal dramatic shedding of the liquid near the nozzle, an effect that has been predicted, but to our knowledge never before imaged. These shedding structures have an approximate spatial frequency of 10 mm-1 with lengths from 50 to 200 μm. Several parameters are explored including injection pressure, liquid fuel temperature, air temperature and pressure, and fuel type. Resulting trends are summarized with accompanying images.

  4. Investigation of High Pressure, Multi-Hole Diesel Fuel Injection Using High Speed Imaging

    Science.gov (United States)

    Morris, Steven; Eagle, Ethan; Wooldridge, Margaret

    2012-10-01

    Research to experimentally capture and understand transient fuel spray behavior of modern fuel injection systems remains underdeveloped. To this end, a high-pressure diesel common-rail fuel injector was instrumented in a spherical, constant volume combustion chamber to image the early time history of injection of diesel fuel. The research-geometry fuel injector has four holes aligned on a radial plane of the nozzle with hole sizes of 90, 110, 130 and 150 μm in diameter. Fuel was injected into a non-reacting environment with ambient densities of 17.4, 24.0, and 31.8 kg/m3 at fuel rail pressures of 1000, 1500, and 2000 bar. High speed images of fuel injection were taken using backlighting at 100,000 frames per second (100 kfps) and an image processing algorithm. The experimental results are compared with a one-dimensional fuel-spray model that was historically developed and applied to fuel sprays from single-hole fuel injectors. Fuel spray penetration distance was evaluated as a function of time for the different injector hole diameters, fuel injection pressures and ambient densities. The results show the differences in model predictions and experimental data at early times in the spray development.

  5. Evaluation of High Pressure Components of Fuel Injection Systems Using Speckle Interferometry

    OpenAIRE

    Basara, Adis

    2007-01-01

    The modern high pressure fuel injection systems installed in engines provide a highly efficient combustion process accompanied by low emissions of exhaust gases and an impressive level of dynamic response. The design and development of mechanical components for such systems pose a great challenge, since they have to operate under extremely high fluctuating pressures (e.g. up to 2000 bar) for a long lifetime (more than 1000 injections per minute). The permanent change between a higher and a lo...

  6. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  7. An experimental study on the effects of high-pressure and multiple injection strategies on DI diesel engine emissions

    KAUST Repository

    Yang, Seung Yeon; Chung, Suk-Ho

    2013-01-01

    An experimental study on effects of high-pressure injections in conjunction with split fuel injections were conducted on an AVL single cylinder DI diesel engine. Various injection schemes were studied through the use of an electronically controlled

  8. An experimental study on the effects of high-pressure and multiple injection strategies on DI diesel engine emissions

    KAUST Repository

    Yang, Seung Yeon

    2013-03-25

    An experimental study on effects of high-pressure injections in conjunction with split fuel injections were conducted on an AVL single cylinder DI diesel engine. Various injection schemes were studied through the use of an electronically controlled, common rail injection system capable of injection pressures up to 200 MPa and a maximum of six injections per combustion event. Up to 100 MPa of the fuel injection pressure, the higher injection pressures create faster combustion rates that result in the higher in-cylinder gas temperatures as compared to conventional low-pressure fuel injection systems. When applying high-pressure injections, particulate emission reductions of up to 50% were observed with no change in hydrocarbon emissions, reductions of CO emissions and only slightly higher NOx emissions. Over 100 MPa, on the other hand, the higher injection pressures still reduced up to almost zero-level of particulate emission, at the same time that the NO emission is reduced greatly. Under these high-pressure injection conditions, strong correlations between soot and CO emissions were observed, which compete for the oxidizing OH species. Multiple or split high-pressure injections also investigated as a means to decrease particulate emissions. As a result, a four-split injection strategy resulted in a 55% reduction in particulates and with little or no penalty on NOx emissions. The high pressure split injection strategy with EGR was more effective in reducing particulate and CO emissions simultaneously. Copyright © 2013 SAE International and Copyright © 2013 TSAE.

  9. Effect of Fuel Injection and Mixing Characteristics on Pulse-Combustor Performance at High-Pressure

    Science.gov (United States)

    Yungster, Shaye; Paxson, Daniel E.; Perkins, Hugh D.

    2014-01-01

    Recent calculations of pulse-combustors operating at high-pressure conditions produced pressure gains significantly lower than those observed experimentally and computationally at atmospheric conditions. The factors limiting the pressure-gain at high-pressure conditions are identified, and the effects of fuel injection and air mixing characteristics on performance are investigated. New pulse-combustor configurations were developed, and the results show that by suitable changes to the combustor geometry, fuel injection scheme and valve dynamics the performance of the pulse-combustor operating at high-pressure conditions can be increased to levels comparable to those observed at atmospheric conditions. In addition, the new configurations can significantly reduce the levels of NOx emissions. One particular configuration resulted in extremely low levels of NO, producing an emission index much less than one, although at a lower pressure-gain. Calculations at representative cruise conditions demonstrated that pulse-combustors can achieve a high level of performance at such conditions.

  10. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  11. Method of injecting iron ion into reactor coolant

    International Nuclear Information System (INIS)

    Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

    1988-01-01

    Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

  12. Modern high pressure gas injection centrifugal compressor for enhanced oil recovery

    Energy Technology Data Exchange (ETDEWEB)

    Almasi, Amin [Worley Parsons Services Pty Ltd, Brisbane, NSW (Australia). Mechanical Dept.

    2011-12-15

    This article covers different design, manufacturing, performance and reliability aspects of modern high pressure gas re-injection centrifugal compressor units. Advances and recent technologies on critical areas such as rotor dynamics, anti-surge system, rotating stall prevention, auxiliary systems, material selection, shop performance tests and gas sealing are studied. Three different case studies for modern re-injection machines including 12 MW, 15 MW and 32 MW trains are presented. (orig.)

  13. Simulation of a loss of coolant accident with hydroaccumulator injection

    International Nuclear Information System (INIS)

    1988-10-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. The IAEA, having identified the need for experimental data due to the difficulties of building integral test facilities and the high costs of these experiments, has accepted the offer of the Hungarian Academy of Sciences and organized two standard problem exercises. In these exercises, experimental data from the simulation of a 7.4% break loss of coolant accident was compared with analytical prediction of the behaviour of the facility calculated with computer codes. The second standard problem exercise involved a similar test, with the exception that in this case hydroaccumulator of the safety injection system were allowed to inject water in the system as anticipated in the design of the plant. This document presents a complete overview of the Second Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation. 22 refs, figs and tabs

  14. Accident tolerant high-pressure helium injection system concept for light water reactors

    International Nuclear Information System (INIS)

    Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

    2016-01-01

    Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

  15. Precision tubes for high-pressure diesel injection lines; Praezisrohre fuer Hochdruck-Dieseleinspritzleitungen

    Energy Technology Data Exchange (ETDEWEB)

    Hagedorn, M.; Lechtenfeld, U.; Zaremba, A. [Mannesmann Praezisrohr GmbH, Hamm (Germany)

    2008-03-15

    The requirements on diesel injection lines raise because of increasing customers demands and more rigid environmental laws. In this context higher injection pressures effect both aspects positively. One important condition for increasing pressure levels is the economical provision of suitable injection lines. To reach this aim, Mannesmann Praezisrohr GmbH developed precision tubes for injection lines, which are fulfilling these increasing requirements. (orig.)

  16. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Khattab, M; Ibrahim, N A; Bedrose, C D [Reactors department, nuclear research center, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs.

  17. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

    1995-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

  18. Pressure behaviour in a nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    KHattab, M.S.; Ibrahim, N.A.; Bedrose, S.D.

    1994-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break, is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The scenarios of small, medium and large LOCA at 2%, 15% and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The results of large LOCA showed good agreement with westinghouse calculations of the same design. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs., 1 tab

  19. Recent Experimental Efforts on High-Pressure Supercritical Injection for Liquid Rockets and Their Implications

    Directory of Open Access Journals (Sweden)

    Bruce Chehroudi

    2012-01-01

    Full Text Available Pressure and temperature of the liquid rocket thrust chambers into which propellants are injected have been in an ascending trajectory to gain higher specific impulse. It is quite possible then that the thermodynamic condition into which liquid propellants are injected reaches or surpasses the critical point of one or more of the injected fluids. For example, in cryogenic hydrogen/oxygen liquid rocket engines, such as Space Shuttle Main Engine (SSME or Vulcain (Ariane 5, the injected liquid oxygen finds itself in a supercritical condition. Very little detailed information was available on the behavior of liquid jets under such a harsh environment nearly two decades ago. The author had the opportunity to be intimately involved in the evolutionary understanding of injection processes at the Air Force Research Laboratory (AFRL, spanning sub- to supercritical conditions during this period. The information included here attempts to present a coherent summary of experimental achievements pertinent to liquid rockets, focusing only on the injection of nonreacting cryogenic liquids into a high-pressure environment surpassing the critical point of at least one of the propellants. Moreover, some implications of the results acquired under such an environment are offered in the context of the liquid rocket combustion instability problem.

  20. Disruption mitigation with high-pressure helium gas injection on EAST tokamak

    Science.gov (United States)

    Chen, D. L.; Shen, B.; Granetz, R. S.; Qian, J. P.; Zhuang, H. D.; Zeng, L.; Duan, Y.; Shi, T.; Wang, H.; Sun, Y.; Xiao, B. J.

    2018-03-01

    High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time  ⩽150 μs, capable of injecting up to 7 × 1022 particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas (N_He≃ N_plasma ) can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity (N_He≃ 40× N_plasma ) can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (˜7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas.

  1. Recommended HPI [High Pressure Injection] rates for the TMI-2 analysis exercise (0 to 300 minutes)

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1987-09-01

    An international analysis exercise has been organized to evaluate the ability of nuclear reactor severe accident computer codes to predict the TMI-2 accident sequence and core damage progression during the first 300 minutes of the accident. A required boundary condition for the analysis exercise is the High Pressure Injection or make-up rates into the primary system during the accident. Recommended injection rates for the first 300 minutes of the accident are presented. Recommendations for several sensitivity studies are also presented. 6 refs., 5 figs., 1 tab

  2. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  3. Modelling of the work processes high-pressure pump of common rail diesel injection system

    Directory of Open Access Journals (Sweden)

    Botwinska Katarzyna

    2016-01-01

    Full Text Available Common rail injection systems are becoming a more widely used solution in the fuel systems of modern diesel engines. The main component and the characteristic feature of the system is rail injection of the fuel under high pressure, which is passed to the injector and further to the combustion chamber. An important element in this process is the high-pressure pump, continuing adequate pressure in the rail injection system. Common rail (CR systems are being modified in order to optimise their work and virtual simulations are a useful tool in order to analyze the correctness of operation of the system while varying the parameters and settings, without any negative impact on the real object. In one particular study, a computer simulation of the pump high-pressure CR system was made in MatLab environment, based on the actual dimensions of the object – a one-cylinder diesel engine, the Farymann Diesel 18W. The resulting model consists of two parts – the first is responsible for simulating the operation of the high-pressure pump, and the second responsible for simulation of the remaining elements of the CR system. The results of this simulation produced waveforms of the following parameters: fluid flow from the manifold to the injector [m3/s], liquid flow from the manifold to the atmosphere [m3/s], and manifold pressure [Pa]. The simulation results allow for a positive verification of the model and the resulting system could become a useful element of simulation of the entire position and control algorithm.

  4. Physics based simulation of seismicity induced in the vicinity of a high-pressure fluid injection

    Science.gov (United States)

    McCloskey, J.; NicBhloscaidh, M.; Murphy, S.; O'Brien, G. S.; Bean, C. J.

    2013-12-01

    High-pressure fluid injection into subsurface is known, in some cases, to induce earthquakes in the surrounding volume. The increasing importance of ';fracking' as a potential source of hydrocarbons has made the seismic hazard from this effect an important issue the adjudication of planning applications and it is likely that poor understanding of the process will be used as justification of refusal of planning in Ireland and the UK. Here we attempt to understand some of the physical controls on the size and frequency of induced earthquakes using a physics-based simulation of the process and examine resulting earthquake catalogues The driver for seismicity in our simulations is identical to that used in the paper by Murphy et al. in this session. Fluid injection is simulated using pore fluid movement throughout a permeable layer from a high-pressure point source using a lattice Boltzmann scheme. Diffusivities and frictional parameters can be defined independently at individual nodes/cells allowing us to reproduce 3-D geological structures. Active faults in the model follow a fractal size distribution and exhibit characteristic event size, resulting in a power-law frequency-size distribution. The fluid injection is not hydraulically connected to the fault (i.e. fluid does not come into physical contact with the fault); however stress perturbations from the injection drive the seismicity model. The duration and pressure-time function of the fluid injection can be adjusted to model any given injection scenario and the rate of induced seismicity is controlled by the local structures and ambient stress field as well as by the stress perturbations resulting from the fluid injection. Results from the rate and state fault models of Murphy et al. are incorporated to include the effect of fault strengthening in seismically quite areas. Initial results show similarities with observed induced seismic catalogues. Seismicity is only induced where the active faults have not been

  5. Injection halos of hydrocarbons above oil-gas fields with super-high pressures

    Energy Technology Data Exchange (ETDEWEB)

    Bakhtin, V.V.

    1979-09-01

    We studied the origin of injection halos of hydrocarbons above oil-gas fields with anomalously high formation pressures (AHFP). Using fields in Azerbaydzhan and Chechen-Ingushetiya as an example, we demonstrate the effect of certain factors (in particular, faults and zones of increased macro- and micro-jointing) on the morpholoy of the halos. The intensity of micro-jointing (jointing permeability, three-dimensional density of micro-jointing) is directly connected with vertical dimensions of the halos. We measured halos based on transverse profiles across the Khayan-Kort field and studied the distribution of bitumen saturation within the injection halo. Discovery of injection halos during drilling has enabled us to improve the technology of wiring deep-seated exploratory wells for oil and gas in regions with development of AHFP.

  6. Influence of Powder Injection Parameters in High-Pressure Cold Spray

    Science.gov (United States)

    Ozdemir, Ozan C.; Widener, Christian A.

    2017-10-01

    High-pressure cold spray systems are becoming widely accepted for use in the structural repair of surface defects of expensive machinery parts used in industrial and military equipment. The deposition quality of cold spray repairs is typically validated using coupon testing and through destructive analysis of mock-ups or first articles for a defined set of parameters. In order to provide a reliable repair, it is important to not only maintain the same processing parameters, but also to have optimum fixed parameters, such as the particle injection location. This study is intended to provide insight into the sensitivity of the way that the powder is injected upstream of supersonic nozzles in high-pressure cold spray systems and the effects of variations in injection parameters on the nature of the powder particle kinetics. Experimentally validated three-dimensional computational fluid dynamics (3D CFD) models are implemented to study the particle impact conditions for varying powder feeder tube size, powder feeder tube axial misalignment, and radial powder feeder injection location on the particle velocity and the deposition shape of aluminum alloy 6061. Outputs of the models are statistically analyzed to explore the shape of the spray plume distribution and resulting coating buildup.

  7. PTV analysis of the entrained air into the diesel spray at high-pressure injection

    Science.gov (United States)

    Toda, Naoki; Yamashita, Hayato; Mashida, Makoto

    2014-08-01

    In order to clarify the effect of high-pressure injection on soot reduction in terms of the air entrainment into spray, the air flow surrounding the spray and set-off length indicating the distance from the nozzle tip to the flame region in diffusion diesel combustion were investigated using 300MPa injection of a multi-hole injector. The measurement of the air entrainment flow was carried out at non-evaporating condition using consecutive PTV (particle tracking velocimetry) method with a high-speed camera and a high-frequency pulse YAG laser. The set-off length was measured at highpressure and high-temperature using the combustion bomb of constant volume and optical system of shadow graph method. And the amount of air entrainment into spray until reaching set-off length in diffusion combustion was studied as a factor of soot formation.

  8. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  9. Direct injection of high pressure gas : scaling properties of pulsed turbulent jets

    NARCIS (Netherlands)

    Baert, R.S.G.; Klaassen, A.; Doosje, E.

    2010-01-01

    Existing gasoline DI injection equipment has been modified to generate single hole pulsed gas jets. Injection experiments have been performed at combinations of 3 different pressure ratios (2 of which supercritical) respectively 3 different hole geometries (i.e. length to diameter ratios). Injection

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

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

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

  11. Experimental and analytical study on biodiesel and diesel spray characteristics under ultra-high injection pressure

    International Nuclear Information System (INIS)

    Wang Xiangang; Huang Zuohua; Kuti, Olawole Abiola; Zhang Wu; Nishida, Keiya

    2010-01-01

    Spray characteristics of biodiesels (from palm and cooked oil) and diesel under ultra-high injection pressures up to 300 MPa were studied experimentally and analytically. Injection delay, spray penetration, spray angle, spray projected area and spray volume were measured in a spray vessel using a high speed video camera. Air entrainment and atomization characteristics were analyzed with the quasi-steady jet theory and an atomization model respectively. The study shows that biodiesels give longer injection delay and spray tip penetration. Spray angle, projected area and volume of biodiesels are smaller than those of diesel fuel. The approximately linear relationship of non-dimensional spray tip penetration versus time suggests that the behavior of biodiesel and diesel sprays is similar to that of gaseous turbulent jets. Calculation from the quasi-steady jet theory shows that the air entrainment of palm oil is worse than that of diesel, while the cooked oil and diesel present comparable air entrainment characteristics. The estimation on spray droplet size shows that biodiesels generate larger Sauter mean diameter due to higher viscosity and surface tension.

  12. Energy efficiency of a direct-injection internal combustion engine with high-pressure methanol steam reforming

    International Nuclear Information System (INIS)

    Poran, Arnon; Tartakovsky, Leonid

    2015-01-01

    This article discusses the concept of a direct-injection ICE (internal combustion engine) with thermo-chemical recuperation realized through SRM (steam reforming of methanol). It is shown that the energy required to compress the reformate gas prior to its injection into the cylinder is substantial and has to be accounted for. Results of the analysis prove that the method of reformate direct-injection is unviable when the reforming is carried-out under atmospheric pressure. To reduce the energy penalty resulted from the gas compression, it is suggested to implement a high-pressure reforming process. Effects of the injection timing and the injector's flow area on the ICE-SRM system's fuel conversion efficiency are studied. The significance of cooling the reforming products prior to their injection into the engine-cylinder is demonstrated. We show that a direct-injection ICE with high-pressure SRM is feasible and provides a potential for significant efficiency improvement. Development of injectors with greater flow area shall contribute to further efficiency improvements. - Highlights: • Energy needed to compress the reformate is substantial and has to be accounted for. • Reformate direct-injection is unviable if reforming is done at atmospheric pressure. • Direct-injection engine with high-pressure methanol reforming is feasible. • Efficiency improvement by 12–14% compared with a gasoline-fed engine was shown

  13. Opportunities and challenges in green house gases reduction using high pressure direct injection of natural gas

    International Nuclear Information System (INIS)

    Ouellette, P.

    2001-01-01

    In an effort to reduce Greenhouse Gases, Westport Innovations is developing a high pressure direct injection (HPDI) technology for gaseous fuels. This technology adapts the diesel cycle for gaseous fuels, since the diesel cycle provides high efficiency, high low-speed torque, fast transient capabilities and reliability. Because of their high efficiency, diesels are very favorable from a Greenhouse Gas (GHG) point of view, however they remain challenged by high nitrogen oxides (NOx) and particulate matter (PM) emissions. When directly injecting natural gas, NOx and PM emissions can be reduced by approximately 50% while maintaining the performance of the diesel engine. This allows the use of abundant and historically cheaper natural gas. Because of its lower carbon content per unit energy, natural gas also offers further GHG reduction over the diesel if the efficiency is preserved and if methane emissions are low. This paper discusses development efforts at Westport for several applications including on-highway trucks, light-duty delivery trucks and power generation

  14. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

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

  15. Construction of a Direct Water-Injected Two-Stroke Engine for Phased Direct Fuel Injection-High Pressure Charging Investigations

    Science.gov (United States)

    Somsel, James P.

    1998-01-01

    The development of a water injected Orbital Combustion Process (OCP) engine was conducted to assess the viability of using the powerplant for high altitude NASA aircraft and General Aviation (GA) applications. An OCP direct fuel injected, 1.2 liter, three cylinder, two-stroke engine has been enhanced to independently inject water directly into the combustion chamber. The engine currently demonstrates low brake specific fuel consumption capability and an excellent power to weight ratio. With direct water injection, significant improvements can be made to engine power, to knock limits/ignition advance timing, and to engine NO(x) emissions. The principal aim of the testing was to validate a cyclic model developed by the Systems Analysis Branch at NASA Ames Research Center. The work is a continuation of Ames' investigations into a Phased Direct Fuel Injection Engine with High Pressure Charging (PDFI-ITPC).

  16. The study of two methods for high pressure injection in CT enhancement to display the aortic dissecting aneurysm

    International Nuclear Information System (INIS)

    Wang Yang; Zhu Bin; Zhang Zhen

    2008-01-01

    Objective: To discuss the consequences of two different methods of high pressure injection in CT contrast enhancement to display the aortic dissecting aneurysm. Methods: 100 patients underwent Lightspeed 16 MS CT with contrast enhancement of Stellant D high pressure injector (Medrad), injecting speed of 4.0 mL/s and 80 ml dosage of contrast medium. All patients were divided into A and B groups with 50 in each. The single high pressure injection was applied to A group without isotonic Na chloride flush. B group underwent the same high pressure injection and followed by isotonic Na chloride flush. The method of evaluation was carried out by double blind observation. Results: A group revealed radiologic artifact up to 40 cases with positive rate of 80%. B group demonstrated the same kind of radiologic artifact in 26 cases with positive rate of 52%. Conclusions: Using normal saline (sodium chloride solution)flush after high pressure injection of contrast medium during MSCT angiography is obviously better to demonstrate the internal structures of treatvessels. (authors)

  17. Simulation of coolant mixing in pressure vessel reactors

    International Nuclear Information System (INIS)

    Hoehne, T.

    2003-06-01

    The work was aimed at the experimental investigation and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing conductivity measurements by wire mesh sensors and velocity measurements by the LDA technique. The CFD calculations were carried out with the CFD-code CFX-4. For the design of the facility, calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. Flow measurements and the corresponding CFD calculations in the ROCOM downcomer under steady state conditions showed a Re number independency at nominal flow rates. The flow field is dominated by recirculation areas below the inlet nozzles. Transient flow measurements with high performance LDA-technique showed in agreement with CFX-4 results, that in the case of the start up of a pump after a laminar stage large vortices dominate the flow. In the case of stationary mixing, the maximum value of the averaged mixing scalar at the core inlet was found in the sector below the inlet nozzle, where the tracer was injected. At the start-up case of one pump due to a strong impulse driven flow at the inlet nozzle the horizontal part of the flow dominates in the downcomer. The injection is distributed into two main jets, the maximum of the tracer concentration at the core inlet appears at the opposite part of the loop where the tracer was injected. Additionally, the stationary three-dimensional flow distribution in the downcomer and the lower plenum of a VVER-440/V-230 reactor was calculated with CFX-4. The comparison with experimental data and an analytical mixing model showed a

  18. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  19. Partial Discharge Measurements in HV Rotating Machines in Dependence on Pressure of Coolant

    Directory of Open Access Journals (Sweden)

    I. Kršňák

    2002-01-01

    Full Text Available The influence of the pressure of the coolant used in high voltage rotating machines on partial discharges occurring in stator insulation is discussed in this paper. The first part deals with a theoretical analysis of the topic. The second part deals with the results obtained on a real generator in industrial conditions. Finally, theoretical assumptions and obtained results are compared.

  20. Analysis of large break loss of coolant accident with simultaneous injection into cold leg and hot leg

    International Nuclear Information System (INIS)

    Luo Bangqi

    1997-01-01

    When a large break loss of coolant accident occurs, the most part of the safety injection water injected into the cold leg by the safety injection system will flow through the channel between the pressure vessel and the barrel out of the break into the containment, only a little part of the safety injection water can flow into the reactor core. If the safety injection can inject into both the cold leg and the hot leg simultaneously, the safety injection water injected from the cold leg will flow into the core more easily, because the safety injection water injected from the hot leg will carry out more heat from the upper plenum and the core, so the upper plenum and the core is depressed. In addition, a small part of the safety injection water injected from the hot leg will flow down in the core after impinging the guide tubes in the upper plenum, so the core will get more safety injection water than only cold leg injection, and the core will be much safer

  1. Evaluation of Coolant Injection Procedure in the Severe Accident Management Strategy of APR1400

    International Nuclear Information System (INIS)

    Cho, Yongjin; Lim, Kukhee; Song, Sungchu; Lee, Sukho; Hwang, Taesuk

    2013-01-01

    A coolant injection strategy in the severe accident management guideline (SAMG) of APR1400 relates to immediate coolant injection into RCS (Reactor Coolant System) or injection following the recovery of secondary coolant inventory. This strategy could play important role in accident mitigation and radiological consequences. In this study, appropriateness of the strategy was evaluated using MELCOR1.8.6 and several sensitivity studies of the key parameters were performed. Analysis for APR1400 using MELCOR 1.8.6 was performed to evaluate the effectiveness of accident management strategies and the following conclusions were identified. Sequential operation of secondary and RCS injection may not be the best strategy and the simultaneous injection of secondary and RCS injection could be more preferable. At least, the RCS injection should start before complete drainage of water in the safety injection tank using mobile pumps. In this study, the effectiveness of timing of operator action has been examined and the amount of injection flowrate needs to be studied in the future

  2. Advanced Production Surface Preparation Technology Development for Ultra-High Pressure Diesel Injection

    Energy Technology Data Exchange (ETDEWEB)

    Grant, Marion B.

    2012-04-30

    In 2007, An Ultra High Injection Pressure (UHIP) fueling method has been demonstrated by Caterpillar Fuel Systems - Product Development, demonstrating ability to deliver U.S. Environment Protection Agency (EPA) Tier 4 Final diesel engine emission performance with greatly reduced emissions handling components on the engine, such as without NOx reduction after-treatment and with only a through-flow 50% effective diesel particulate trap (DPT). They have shown this capability using multiple multi-cylinder engine tests of an Ultra High Pressure Common Rail (UHPCR) fuel system with higher than traditional levels of CEGR and an advanced injector nozzle design. The system delivered better atomization of the fuel, for more complete burn, to greatly reduce diesel particulates, while CEGR or high efficiency NOx reduction after-treatment handles the NOx. With the reduced back pressure of a traditional DPT, and with the more complete fuel burn, the system reduced levels of fuel consumption by 2.4% for similar delivery of torque and horsepower over the best Tier 4 Interim levels of fuel consumption in the diesel power industry. The challenge is to manufacture the components in high-volume production that can withstand the required higher pressure injection. Production processes must be developed to increase the toughness of the injector steel to withstand the UHIP pulsations and generate near perfect form and finish in the sub-millimeter size geometries within the injector. This project resulted in two developments in 2011. The first development was a process and a machine specification by which a high target of compressive residual stress (CRS) can be consistently imparted to key surfaces of the fuel system to increase the toughness of the steel, and a demonstration of the feasibility of further refinement of the process for use in volume production. The second development was the demonstration of the feasibility of a process for imparting near perfect, durable geometry to

  3. REMIX: a computer program for temperature transients due to high pressure injection after interruption of natural circulation

    International Nuclear Information System (INIS)

    Iyer, K.; Nourbakhsh, H.P.; Theofanous, T.G.

    1986-05-01

    This report describes the features and use of several computer programs developed on the basis of the Regional Mixing Model (RMM). This model provides a phenomenologically-based analytical description of the stratified flow and temperature fields resulting from High Pressure Safety Injection (HPI) in the stagnated loops of a Pressurized Water Reactor (PWR). The basic program is called REMIX and is intended for thermally-induced stratification at low Froude number injections. The REMIX-S version is intended for solute-induced stratification with or without thermal effects as found in several experimental simulations. The NEWMIX program is a derivative of REMIX representing the limit of maximum possible mixing within the cold leg and is intended for high Froude number injections. The NEWMIX-S version accounts for solute effects. Listings of all programs and sample problem input and output files are included. 10 refs

  4. Quantitative Imaging of Turbulent Mixing Dynamics in High-Pressure Fuel Injection to Enable Predictive Simulations of Engine Combustion

    Energy Technology Data Exchange (ETDEWEB)

    Frank, Jonathan H. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Reacting Flows Dept.; Pickett, Lyle M. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.; Bisson, Scott E. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Remote Sensing and Energetic Materials Dept.; Patterson, Brian D. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). combustion Chemistry Dept.; Ruggles, Adam J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Reacting Flows Dept.; Skeen, Scott A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.; Manin, Julien Luc [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.; Huang, Erxiong [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Reacting Flows Dept.; Cicone, Dave J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.; Sphicas, Panos [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Engine Combustion Dept.

    2015-09-01

    In this LDRD project, we developed a capability for quantitative high - speed imaging measurements of high - pressure fuel injection dynamics to advance understanding of turbulent mixing in transcritical flows, ignition, and flame stabilization mechanisms, and to provide e ssential validation data for developing predictive tools for engine combustion simulations. Advanced, fuel - efficient engine technologies rely on fuel injection into a high - pressure, high - temperature environment for mixture preparation and com bustion. Howe ver, the dynamics of fuel injection are not well understood and pose significant experimental and modeling challenges. To address the need for quantitative high - speed measurements, we developed a Nd:YAG laser that provides a 5ms burst of pulses at 100 kHz o n a robust mobile platform . Using this laser, we demonstrated s patially and temporally resolved Rayleigh scattering imaging and particle image velocimetry measurements of turbulent mixing in high - pressure gas - phase flows and vaporizing sprays . Quantitativ e interpretation of high - pressure measurements was advanced by reducing and correcting interferences and imaging artifacts.

  5. Compressed air injection technique to standardize block injection pressures.

    Science.gov (United States)

    Tsui, Ban C H; Li, Lisa X Y; Pillay, Jennifer J

    2006-11-01

    Presently, no standardized technique exists to monitor injection pressures during peripheral nerve blocks. Our objective was to determine if a compressed air injection technique, using an in vitro model based on Boyle's law and typical regional anesthesia equipment, could consistently maintain injection pressures below a 1293 mmHg level associated with clinically significant nerve injury. Injection pressures for 20 and 30 mL syringes with various needle sizes (18G, 20G, 21G, 22G, and 24G) were measured in a closed system. A set volume of air was aspirated into a saline-filled syringe and then compressed and maintained at various percentages while pressure was measured. The needle was inserted into the injection port of a pressure sensor, which had attached extension tubing with an injection plug clamped "off". Using linear regression with all data points, the pressure value and 99% confidence interval (CI) at 50% air compression was estimated. The linearity of Boyle's law was demonstrated with a high correlation, r = 0.99, and a slope of 0.984 (99% CI: 0.967-1.001). The net pressure generated at 50% compression was estimated as 744.8 mmHg, with the 99% CI between 729.6 and 760.0 mmHg. The various syringe/needle combinations had similar results. By creating and maintaining syringe air compression at 50% or less, injection pressures will be substantially below the 1293 mmHg threshold considered to be an associated risk factor for clinically significant nerve injury. This technique may allow simple, real-time and objective monitoring during local anesthetic injections while inherently reducing injection speed.

  6. Development and validation of a model for high pressure liquid poison injection for CANDU-6 shutdown system no.2

    International Nuclear Information System (INIS)

    Rhee, B.-W.; Jeong, C.J.; Choi, J.H.; Yoo, S.-Y.

    2002-01-01

    In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the calandria tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, an AEA Technology CFD code, to simulate the formation and growth of the poison jet curtain inside the moderator tank. For validation, the current model is validated against a poison injection experiment performed at BARC, India and another poison jet experiment for Generic CANDU-6 performed at AECL, Canada. In conclusion this set of models is considered to predict the experimental results in a physically reasonable and consistent manner. (author)

  7. Emulation study on system characteristic of high pressure common-rail fuel injection system for marine medium-speed diesel engine

    Science.gov (United States)

    Wang, Qinpeng; Yang, Jianguo; Xin, Dong; He, Yuhai; Yu, Yonghua

    2018-05-01

    In this paper, based on the characteristic analyzing of the mechanical fuel injection system for the marine medium-speed diesel engine, a sectional high-pressure common rail fuel injection system is designed, rated condition rail pressure of which is 160MPa. The system simulation model is built and the performance of the high pressure common rail fuel injection system is analyzed, research results provide the technical foundation for the system engineering development.

  8. LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressure injection System (HPIS)

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The sixth OECD LOFT experiment was conducted on 5 March 1984. It simulated a 1.8-in cold leg break LOCA with no HPIS available. This experiment was designed mainly for investigation of plant recovery effectiveness using secondary bleed and feed during core uncover and addressed accumulator injection at low pressure differentials. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  9. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    International Nuclear Information System (INIS)

    Daigo Tsuru; Mikio Enoeda; Masato Akiba

    2006-01-01

    Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

  10. Safety assessment of the SMART design during SBLOCA tests using the high pressure safety injection pump of the SMART-ITL facility

    International Nuclear Information System (INIS)

    Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yoon, Eun-Koo; Shin, Yong-Cheol; Min, Kyoung-Ho; Park, Jong-Kuk; Choi, Nam-Hyun; Bang, Yun-Gon; Seo, Chan-Jong; Yi, Sung-Jae; Park, Hyun-Sik

    2016-01-01

    SMART is a small-sized integral pressurized light water reactor designed by the Korea Atomic Energy Research Institute (KAERI) from 1997 and received standard design approval (SDA) by the Korean regulatory body in July 2012. Single reactor pressure vessel contains all of the main components including a pressurizer (PZR), steam generators (SG) and reactor coolant pumps (RCP) without any large-size pipes. Several tests to verify a safety and performance of SMART design were carried out. This paper introduces a comparison with three SBLOCA tests. Overall thermal-hydraulic phenomena were observed and showed a traditional trend to decrease a system pressure and temperature. A collapsed water level of the hot side indicated that the safety injection system was successfully operated to recover the reactor coolant system (RCS) and protect the core uncover. An SBLOCA test simulating a guillotine break on the SIS, SCS, and PSV was performed. It was enough to keep a steady-state condition before the SBLOCA test begins. An actuation signal as the boundary condition was properly simulated during the transient test. The scenarios of the SBLOCA in the SMART design were reproduced well using the SMART-ITL facility. The safety injection is effective to protect the core uncover as well as to cool down the RCS. All of the measured parameters show reasonable behaviors

  11. Safety assessment of the SMART design during SBLOCA tests using the high pressure safety injection pump of the SMART-ITL facility

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yoon, Eun-Koo; Shin, Yong-Cheol; Min, Kyoung-Ho; Park, Jong-Kuk; Choi, Nam-Hyun; Bang, Yun-Gon; Seo, Chan-Jong; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    SMART is a small-sized integral pressurized light water reactor designed by the Korea Atomic Energy Research Institute (KAERI) from 1997 and received standard design approval (SDA) by the Korean regulatory body in July 2012. Single reactor pressure vessel contains all of the main components including a pressurizer (PZR), steam generators (SG) and reactor coolant pumps (RCP) without any large-size pipes. Several tests to verify a safety and performance of SMART design were carried out. This paper introduces a comparison with three SBLOCA tests. Overall thermal-hydraulic phenomena were observed and showed a traditional trend to decrease a system pressure and temperature. A collapsed water level of the hot side indicated that the safety injection system was successfully operated to recover the reactor coolant system (RCS) and protect the core uncover. An SBLOCA test simulating a guillotine break on the SIS, SCS, and PSV was performed. It was enough to keep a steady-state condition before the SBLOCA test begins. An actuation signal as the boundary condition was properly simulated during the transient test. The scenarios of the SBLOCA in the SMART design were reproduced well using the SMART-ITL facility. The safety injection is effective to protect the core uncover as well as to cool down the RCS. All of the measured parameters show reasonable behaviors.

  12. Study on primary coolant system depressurization effect factor in pressurized water reactor

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

    The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

  13. Analysis of a water-coolant leak into a very high-temperature vitrification chamber

    International Nuclear Information System (INIS)

    Felicione, F. S.

    1998-01-01

    A coolant-leakage incident occurred during non-radioactive operation of the Plasma Hearth Process waste-vitrification development system at Argonne National Laboratory when a stray electric arc ruptured az water-cooling jacket. Rapid evaporation of the coolant that entered the very high-temperature chamber pressurized the normally sub-atmospheric system above ambient pressure for over 13 minutes. Any positive pressurization, and particularly a lengthy one, is a safety concern since this can cause leakage of contaminants from the system. A model of the thermal phenomena that describe coolant/hot-material interactions was developed to better understand the characteristics of this type of incident. The model is described and results for a variety of hypothetical coolant-leak incidents are presented. It is shown that coolant leak rates above a certain threshold will cause coolant to accumulate in the chamber, and evaporation from this pool can maintain positive pressure in the system long after the leak has been stopped. Application of the model resulted in reasonably good agreement with the duration of the pressure measured during the incident. A closed-form analytic solution is shown to be applicable to the initial leak period in which the peak pressures are generated, and is presented and discussed

  14. Integrated main coolant pumps for pressurized-water reactors

    International Nuclear Information System (INIS)

    Wieser, R.

    1975-01-01

    The efficiency of an integrated main coolant pump for PWR's is increased. For this purpose, the pump is installed eccentric relative to the vertical axis of the U-type steam generator in the three-section HP chamber in such a way that its impeller wheel and the shell of the latter penetrate into the outlet chamber. The axis of the pump lies in the vertical plane of symmetry of the outlet chamber of the steam generator. The suction tube is arranged in the outlet chamber. To allow it to be installed, it is manufactured out of several parts. The diffusor tube, which is also made of several components, is attached to the horizontal separation plate between the outlet chamber and the pressure chamber so as to penetrate into it. To improve the outflow conditions at the diffusor tube, a plowshare-shaped baffle shield is installed between the diffusor tube and the HP chamber. Moreover, in order to improve the outflow conditions from the pump and from the pressure chamber, the outflow opening of the pressure chamber is put into the cylindrical shell of the HP chamber. In this way, the tensioning anchor is located between the pump and the outlet opening. (DG/RF) [de

  15. Analysis Of Primary Coolant Suction Side Pressure In The Delay Chamber Of The RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto

    2000-01-01

    Delay chamber is a tank to delay flow that located in the primary cooling suction side of RSG-GAS. A void occurred when operation reactor caused by too high the delta P at inlet suction pump. The condition may be avoided by using one line mode of the cooling flow. The analysis show that void volume in the delay chamber is occurred because the coolant negative pressure lowers the saturation pressure should be avoided though decreasing the delta P until about 0.1 bar at about 45 exp 0 C. Solution suggested are to use bypass flow from the spent fuel to the delay chamber. Coolant temperature can be also decreased by decreasing the power level of the reactor as well as improving the heat exchanger and cooling tower performances

  16. Bandwidth of reactor internals vibration resonance with coolant pressure oscillations

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    In a few decades a significant increase in a part of an electricity development on the NPP will require NPP to be operated in non full capacity modes and increase in operation time in transitive modes. Operating in such conditions as compared to the operation on a constant mode will lead to the increase in cyclic dynamical loading. In water cooled water moderated reactors these loading are realized as low-cyclic and high-cyclic loadings. High-cyclic loadings increases are caused by a raised vibration in non stationary modes of operation. It is known, that in some modes of a non full capacity reactor high-cyclic dynamic loadings can increase. It is obvious, that the development of management technologies is necessary for the life time management operation. In the context of this problem one of the main tasks are revealing and the prevention of the conditions of the occurrence of the operation leading to the resonant interaction of the coolant fluctuations and the equipment, reactor vessel (RV), fuel assemblies (FA) and reactor internals (RI) vibration. To prevent the appearance of the conditions for resonance interaction between the fluid flow and the equipments, it is necessary to provide the different frequencies for the self oscillations in the separated elements of the circulating system and also in the parts of the system formed by the comprising of these elements. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of coolant outside of which there is no resonant interaction. The presented work is devoted to finding the solution of this problem. There are results of theoretical an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. The accordance of results had been calculated with had been measured are satisfied for practical purposes. These

  17. Analysis of containment pressure and temperature changes following loss of coolant accident (LOCA)

    International Nuclear Information System (INIS)

    Nguyen Van Thai; Kieu Ngoc Dung

    2015-01-01

    This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories. (author)

  18. Impact of CO2 injection protocol on fluid-solid reactivity: high-pressure and temperature microfluidic experiments in limestone

    Science.gov (United States)

    Jimenez-Martinez, Joaquin; Porter, Mark; Carey, James; Guthrie, George; Viswanathan, Hari

    2017-04-01

    Geological sequestration of CO2 has been proposed in the last decades as a technology to reduce greenhouse gas emissions to the atmosphere and mitigate the global climate change. However, some questions such as the impact of the protocol of CO2 injection on the fluid-solid reactivity remain open. In our experiments, two different protocols of injection are compared at the same conditions (8.4 MPa and 45 C, and constant flow rate 0.06 ml/min): i) single phase injection, i.e., CO2-saturated brine; and ii) simultaneous injection of CO2-saturated brine and scCO2. For that purpose, we combine a unique high-pressure/temperature microfluidics experimental system, which allows reproducing geological reservoir conditions in geo-material substrates (i.e., limestone, Cisco Formation, Texas, US) and high resolution optical profilometry. Single and multiphase flow through etched fracture networks were optically recorded with a microscope, while processes of dissolution-precipitation in the etched channels were quantified by comparison of the initial and final topology of the limestone micromodels. Changes in hydraulic conductivity were quantified from pressure difference along the micromodel. The simultaneous injection of CO2-saturated brine and scCO2, reduced the brine-limestone contact area and also created a highly heterogeneous velocity field (i.e., low velocities regions or stagnation zones, and high velocity regions or preferential paths), reducing rock dissolution and enhancing calcite precipitation. The results illustrate the contrasting effects of single and multiphase flow on chemical reactivity and suggest that multiphase flow by isolating parts of the flow system can enhance CO2 mineralization.

  19. The Schoonebeek Oilfield: the Rw-2e High Pressure Steam Injection Project Gisement de Schoonebeek : le projet RW-2E d'injection de vapeur à haute pression

    Directory of Open Access Journals (Sweden)

    Holtam V. R.

    2006-11-01

    Full Text Available The daily oil production from the Schoonebeek Oilfield amounts to some 1400 m3 /d, of which ca. 65% is produced from a high pressure (85 bar steam injection project. This project was started in 1981 and originally consisted of 7 structurally downdip/middip steam injectors. However, following the initially somewhat disappointing project performance, steam injection was moved to 4 middip/ updip injectors in 1984. This change in the location of the steam injectors, together with an increase in the level of surveillance and a more pragmatic reservoir management policy, has resulted in improved project performance. The ultimate extra oil/steam ratio for the total project is now expected to be 0. 7 m3 oil/ton of steam injected. La production de pétrole du gisement de Schoonebeek est d'environ 1400 m3/jour, dont près de 65% sont obtenus par injection de vapeur à haute pression (85 bar. Ce projet lancé en 1981 comportait initialement 7 injecteurs de vapeur orientés vers l'aval-pendage. En raison de performances décevantes, l'injection de vapeur a été transférée en 1984 sur 4 injecteurs travaillant vers l'amont-pendage. Ce changement de position des injecteurs, accompagné d'une surveillance renforcée et d'une politique de gestion du gisement plus pragmatique, a donné des résultats favorables. On pense que le rapport pétrole/vapeur pour l'ensemble du projet devrait être en dernière analyse de 0,7 m3 de pétrole par tonne de vapeur injectée.

  20. Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

  1. The corrosion products in the coolant circuits of pressurized water nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

  2. Microbially Enhanced Oil Recovery by Sequential Injection of Light Hydrocarbon and Nitrate in Low- And High-Pressure Bioreactors.

    Science.gov (United States)

    Gassara, Fatma; Suri, Navreet; Stanislav, Paul; Voordouw, Gerrit

    2015-10-20

    Microbially enhanced oil recovery (MEOR) often involves injection of aqueous molasses and nitrate to stimulate resident or introduced bacteria. Use of light oil components like toluene, as electron donor for nitrate-reducing bacteria (NRB), offers advantages but at 1-2 mM toluene is limiting in many heavy oils. Because addition of toluene to the oil increased reduction of nitrate by NRB, we propose an MEOR technology, in which water amended with light hydrocarbon below the solubility limit (5.6 mM for toluene) is injected to improve the nitrate reduction capacity of the oil along the water flow path, followed by injection of nitrate, other nutrients (e.g., phosphate) and a consortium of NRB, if necessary. Hydrocarbon- and nitrate-mediated MEOR was tested in low- and high-pressure, water-wet sandpack bioreactors with 0.5 pore volumes of residual oil in place (ROIP). Compared to control bioreactors, those with 11-12 mM of toluene in the oil (gained by direct addition or by aqueous injection) and 80 mM of nitrate in the aqueous phase produced 16.5 ± 4.4% of additional ROIP (N = 10). Because toluene is a cheap commodity chemical, HN-MEOR has the potential to be a cost-effective method for additional oil production even in the current low oil price environment.

  3. Injection characteristics study of high-pressure direct injector for Compressed Natural Gas (CNG) using experimental and analytical method

    Science.gov (United States)

    Taha, Z.; Rahim, MF Abdul; Mamat, R.

    2017-10-01

    The injection characteristics of direct injector affect the mixture formation and combustion processes. In addition, the injector is converted from gasoline operation for CNG application. Thus measurement of CNG direct injector mass flow rate was done by independently tested a single injector on a test bench. The first case investigated the effect of CNG injection pressure and the second case evaluate the effect of pulse-width of injection duration. An analytical model was also developed to predict the mass flow rate of the injector. The injector was operated in a choked condition in both the experiments and simulation studies. In case 1, it was shown that mass flow rate through the injector is affected by injection pressure linearly. Based on the tested injection pressure of 20 bar to 60 bar, the resultant mass flow rate are in the range of 0.4 g/s to 1.2 g/s which are met with theoretical flow rate required by the engine. However, in Case 2, it was demonstrated that the average mass flow rate at short injection durations is lower than recorded in Case 1. At injection pressure of 50 bar, the average mass flow rate for Case 2 and Case 1 are 0.7 g/s and 1.1 g/s respectively. Also, the measured mass flow rate at short injection duration showing a fluctuating data in the range of 0.2 g/s - 1.3 g/s without any noticeable trends. The injector model able to predict the trend of the mass flow rate at different injection pressure but unable to track the fluctuating trend at short injection duration.

  4. Optimal design of Tilting-Pad Thrust Bearings with High Pressure Injection Pockets

    DEFF Research Database (Denmark)

    Heinrichson, Niels; Santos, Ilmar

    2006-01-01

    A thermo-elasto-hydrodynamic(TEHD) model based on the Reynolds equation has been used to study the effect of oil injection pockets on the performance of tilting pad thrust bearings. The optimal position of the pivot both with respect to load carrying capacity and minimal power consumption is seen...

  5. APPLICATION OF MULTIHOLE PRESSURE PROBE FOR RESEARCH OF COOLANT VELOCITY PROFILE IN NUCLEAR REACTOR FUEL ASSEMBLIES

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2015-01-01

    Full Text Available Development of heat and mass transfer intensifiers is a major engineering task in the design of new and modernization of existing fuel assemblies. These devices create lateral mass flow of coolant. Design of intensifiers affects both the coolant mixing and the hydraulic resistance. The aim of this work is to develop a methodology of measuring coolant local velocity in the fuel assembly models with different mixing grids. To solve the problems was manufactured and calibrated multihole pressure probe. The air flow velocity measuring method with multihole pressure probe was used in the experimental studies on the coolant local hydrodynamics in fuel assemblies with mixing grids. Analysis of the coolant lateral velocity vector fields allowed to study the formation of the secondary vortex flows behind the mixing grids, and to determine the basic laws of coolant flow in experimental models. Quantitative data on the coolant flow velocity distribution obtained with a multihole pressure probe make possible to determine the magnitude of the flow lateral velocities in fuel rod gaps, as well as to determine the distance at which damping occurs during mixing. 

  6. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  7. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  8. Analysis of Pressure Pulsation Induced by Rotor-Stator Interaction in Nuclear Reactor Coolant Pump

    Directory of Open Access Journals (Sweden)

    Xu Zhang

    2017-01-01

    Full Text Available The internal flow of reactor coolant pump (RCP is much more complex than the flow of a general mixed-flow pump due to high temperature, high pressure, and large flow rate. The pressure pulsation that is induced by rotor-stator interaction (RSI has significant effects on the performance of pump; therefore, it is necessary to figure out the distribution and propagation characteristics of pressure pulsation in the pump. The study uses CFD method to calculate the behavior of the flow. Results show that the amplitudes of pressure pulsation get the maximum between the rotor and stator, and the dissipation rate of pressure pulsation in impellers passage is larger than that in guide vanes passage. The behavior is associated with the frequency of pressure wave in different regions. The flow rate distribution is influenced by the operating conditions. The study finds that, at nominal flow, the flow rate distribution in guide vanes is relatively uniform and the pressure pulsation amplitude is the smallest. Besides, the vortex shedding or backflow from the impeller blade exit has the same frequency as pressure pulsation but there are phase differences, and it has been confirmed that the absolute value of phase differences reflects the vorticity intensity.

  9. Optical diagnostics of diesel spray injections and combustion in a high-pressure high-temperature cell

    NARCIS (Netherlands)

    Bougie, H.J.T.; Tulej, M.; Dreier, T.; Dam, N.J.; Meulen, J.J. ter; Gerber, T.

    2005-01-01

    We report on spatially and temporally resolved optical diagnostic measurements of propagation and combustion of diesel sprays introduced through a single-hole fuel injector into a constant volume, high-temperature, high-pressure cell. From shadowgraphy images in non-reacting environments of pure

  10. Nonstationary pressure build up in full-pressure containments after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1977-01-01

    The time histories of pressure, temperature and pressure difference during the pressure build up phase of a loss-of-coolant accident (LOCA) in the primary system in full-pressure containments of water cooled nuclear power reactors are treated. These are important for the design of such containments. The experiments within the German research program RS 50 ''Druckverteilung im Containment'' offered, for the first time, the opportunity to observe experimentally fluid-dynamic processes in a multiple divided full-pressure containment, and to test at the same time, computer codes which serve to describe the physical processes during the LOCA. The comparison of the results calculated by the computer codes ZOCO VI and DDIFF with the experimental results showed apparent deviations by special arrangements of the compartments and the vent flow paths of a model containment for the calculation of time dependent pressure-, temperature- and pressure difference-histories. The deviations lead to the development of the analytical model and computer code COFLOW. This new model was primarily designed to deal with the fluid-dynamic processes in the beginning phase of the blowdown as maximal pressure differences appear. Furthermore, it can be used to determine the maximum containment pressure, as well as for long term calculations. The analytical model and computer code COFLOW shows a better correlation between theory and experiment than previous codes

  11. Integrated equipment for increasing and maintaining coolant pressure in primary circuit of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Sykora, D.

    1986-01-01

    An open heat pump circuit is claimed connected to the primary circuit. The pump circuit consists of a steam pressurizer with a built-in steam distributor, a compressor, an expander, a reducing valve, an auxiliary pump, and of water and steam pipes. The operation is described and a block diagram is shown of integrated equipment for increasing and maintaining pressure in the nuclear power plant primary circuit. The appropriate entropy diagram is also shown. The advantage of the open pump circuit consists in reducing the electric power input and electric power consumption for the steam pressurizers, removing entropy loss in heat transfer with high temperature gradient, in the possibility of inserting, between the expander and the auxiliary pump, a primary circuit coolant treatment station, in simplified design and manufacture of the high-pressure steam pressurizer vessel, reducing the weight of the steam pressurizer by changing its shape from cylindrical to spherical, increasing the rate of pressure growth in the primary circuit. (E.S.)

  12. Real-time reactor coolant system pressure/temperature limit system

    International Nuclear Information System (INIS)

    Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

    1991-01-01

    This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

  13. Investigation on transient flow of a centrifugal charging pump in the process of high pressure safety injection

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Fan, E-mail: zhangfan4060@gmail.com; Yuan, Shouqi; Fu, Qiang; Tao, Yi

    2015-11-15

    Highlights: • The transient flow characteristics of the charging pump with the first stage impeller in the HPSI process have been investigated numerically by CFD. • The hydraulic performance of the charging pump during the HPSI are discussed, andthe absolute errors between the simulated and measured results are analyzed in the paper. • Pressure fluctuation in the impeller and flow pattern in the impeller were studied in the HPSI process. It is influenced little at the beginning of the HPSI process while fluctuates strongly in the end of the HPSI process. - Abstract: In order to investigate the transient flow characteristics of the centrifugal charging pump during the transient transition process of high pressure safety injection (HPSI) from Q = 148 m{sup 3}/h to Q = 160 m{sup 3}/h, numerical simulation and experiment are implemented in this study. The transient flow rate, which is the most important factor, is obtained from the experiment and works as the boundary condition to accurately accomplish the numerical simulation in the transient process. Internal characteristics under the variable operating conditions are analyzed through the transient simulation. The results shows that the absolute error between the simulated and measured heads is less than 2.26% and the absolute error between the simulated and measured efficiency is less than 2.04%. Pressure fluctuation in the impeller is less influenced by variable flow rate in the HPSI process, while flow pattern in the impeller is getting better and better with the flow rate increasing. As flow rate increases, fluid blocks on the tongue of the volute and it strikes in this area at large flow rate. Correspondingly, the pressure fluctuation is intense and vortex occurs gradually during this period, which obviously lowers the efficiency of the pump. The contents of the current work can provide references for the design optimization and fluid control of the pump used in the transient process of variable operating

  14. Investigation on transient flow of a centrifugal charging pump in the process of high pressure safety injection

    International Nuclear Information System (INIS)

    Zhang, Fan; Yuan, Shouqi; Fu, Qiang; Tao, Yi

    2015-01-01

    Highlights: • The transient flow characteristics of the charging pump with the first stage impeller in the HPSI process have been investigated numerically by CFD. • The hydraulic performance of the charging pump during the HPSI are discussed, andthe absolute errors between the simulated and measured results are analyzed in the paper. • Pressure fluctuation in the impeller and flow pattern in the impeller were studied in the HPSI process. It is influenced little at the beginning of the HPSI process while fluctuates strongly in the end of the HPSI process. - Abstract: In order to investigate the transient flow characteristics of the centrifugal charging pump during the transient transition process of high pressure safety injection (HPSI) from Q = 148 m"3/h to Q = 160 m"3/h, numerical simulation and experiment are implemented in this study. The transient flow rate, which is the most important factor, is obtained from the experiment and works as the boundary condition to accurately accomplish the numerical simulation in the transient process. Internal characteristics under the variable operating conditions are analyzed through the transient simulation. The results shows that the absolute error between the simulated and measured heads is less than 2.26% and the absolute error between the simulated and measured efficiency is less than 2.04%. Pressure fluctuation in the impeller is less influenced by variable flow rate in the HPSI process, while flow pattern in the impeller is getting better and better with the flow rate increasing. As flow rate increases, fluid blocks on the tongue of the volute and it strikes in this area at large flow rate. Correspondingly, the pressure fluctuation is intense and vortex occurs gradually during this period, which obviously lowers the efficiency of the pump. The contents of the current work can provide references for the design optimization and fluid control of the pump used in the transient process of variable operating conditions.

  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. In-core failure of the instrumented BWR rod by locally induced high coolant temperature

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

  17. Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2017-02-15

    When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

  18. Effect of cross-flow direction of coolant on film cooling effectiveness with one inlet and double outlet hole injection

    Directory of Open Access Journals (Sweden)

    Guangchao Li

    2012-12-01

    Full Text Available In order to study the effect of cross-flow directions of an internal coolant on film cooling performance, the discharge coefficients and film cooling effectiveness with one inlet and double outlet hole injections were simulated. The numerical results show that two different cross-flow directions of the coolant cause the same decrease in the discharge coefficients as that in the case of supplying coolant by a plenum. The different proportion of the mass flow out of the two outlets of the film hole results in different values of the film cooling effectiveness for three different cases of coolant supplies. The film cooling effectiveness is the highest for the case of supplying coolant by the plenum. At a lower blowing ratio of 1.0, the film cooling effectiveness with coolant injection from the right entrance of the passage is higher than that from the left entrance of the passage. At a higher blowing ratio of 2.0, the opposite result is found.

  19. Numerical Analysis on the Influence of Thermal Effects on Oil Flow Characteristic in High-Pressure Air Injection (HPAI Process

    Directory of Open Access Journals (Sweden)

    Hu Jia

    2012-01-01

    Full Text Available In previous laboratory study, we have shown the thermal behavior of Keke Ya light crude oil (Tarim oilfield, branch of CNPC for high-pressure air injection (HPAI application potential study. To clarify the influences of thermal effects on oil production, in this paper, we derived a mathematical model for calculating oil flow rate, which is based on the heat conduction property in porous media from the combustion tube experiment. Based on remarkably limited knowledge consisting of very global balance arguments and disregarding all the details of the mechanisms in the reaction zone, the local governing equations are formulated in a dimensionless form. We use finite difference method to solve this model and address the study by way of qualitative analysis. The time-space dimensionless oil flow rate (qD profiles are established for comprehensive studies on the oil flow rate characteristic affected by thermal effects. It also discusses how these findings will impact HPAI project performances, and several guidelines are suggested.

  20. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  1. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    International Nuclear Information System (INIS)

    Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

    2014-01-01

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

  2. Measurement of gas-liquid two-phase flow around horizontal tube bundle using SF6-water. Simulating high-pressure high-temperature gas-liquid two-phase flow of PWR/SG secondary coolant side at normal pressure

    International Nuclear Information System (INIS)

    Ishikawa, Atsushi; Imai, Ryoj; Tanaka, Takahiro

    2014-01-01

    In order to improve prediction accuracy of analysis code used for design and development of industrial products, technology had been developed to create and evaluate constitutive equation incorporated in analysis code. The experimental facility for PWR/SG U tubes part was manufactured to measure local void fraction and gas-liquid interfacial velocity with forming gas-liquid upward two-phase flow simulating high-pressure high-temperature secondary coolant (water-steam) rising vertically around horizontal tube bundle. The experimental facility could reproduce flow field having gas-liquid density ratio equivalent to real system with no heating using SF6 (Sulfur Hexafluoride) gas at normal temperature and pressure less than 1 MPa, because gas-liquid density ratio, surface tension and gas-liquid viscosity ratio were important parameters to determine state of gas-liquid two-phase flow and gas-liquid density ratio was most influential. Void fraction was measured by two different methods of bi-optical probe and conductivity type probe. Test results of gas-liquid interfacial velocity vs. apparent velocity were in good agreement with existing empirical equation within 10% error, which could confirm integrity of experimental facility and appropriateness of measuring method so as to set up original constitutive equation in the future. (T. Tanaka)

  3. Multi-component vapor-liquid equilibrium model for LES of high-pressure fuel injection and application to ECN Spray A

    NARCIS (Netherlands)

    Matheis, Jan; Hickel, S.

    2018-01-01

    We present and evaluate a two-phase model for Eulerian large-eddy simulations (LES) of liquid-fuel injection and mixing at high pressure. The model is based on cubic equations of state and vapor-liquid equilibrium calculations and can represent the coexistence of supercritical states and

  4. Low pressure injection sequence sensitivity study of the M1 module of MEDICI

    International Nuclear Information System (INIS)

    Corradini, M.L.; Moses, G.A.; Norkus, J.K.; Welzbacker, R.T.

    1985-01-01

    In order to assess the consequences of a PWR containment failure and the ensuing radiological source term following a severe reactor accident, it is necessary to understand the ex-vessel behavior of the molten core. The M1 module of MEDICI models the dynamic fuel-coolant mixing, energetic interaction, and ejection of fuel and coolant from the reactor cavity following such an accident. A sensitivity study of the low pressure injection sequence was performed utilizing a Box-Behnken statistical design to treat five sets of input variables considered to be significant in the mixing and steam explosion processes. The low pressure injection sequence was studied in which the molten corium is modeled as a pour stream entering the cavity without entraining or sweeping out fuel or coolant

  5. The origin and magnitude of pressures in fuel-coolant interactions

    International Nuclear Information System (INIS)

    Heer, W.; Jakeman, D.; Smith, B.L.

    1987-01-01

    A number of small scale experiments to simulate fuel coolant interaction (FCI) effects have been carried out using Freon and water. Contrary to the predictions of most current FCI models, only modest pressure transients are observed within the interaction region itself but large pressure spikes, near to or above critical Freon pressure, are seen at the boundaries of the region. Similar pressure amplification effects have been noticed in parallel experiments involving two phase mixtures. It is suggested that in both cases a water hammer type effect is the cause of the pressure spikes. These observations could form the basis of new thinking in FCI modelling. (author)

  6. Cavity Pressure Behaviour in Micro Injection Moulding

    DEFF Research Database (Denmark)

    Griffiths, C.A.; Dimov, S.S.; Scholz, S.

    2010-01-01

    as well as with the filling of the cavity by the polymer melt. In this paper, two parameters derived from cavity pressure over time (i.e. pressure work). The influence of four µIM parameters (melt temperature, mould temperature, injection speed, aand packing pressure) on the two pressure-related outputs...... has been investigated by moulding a micro fluidic component on three different polymers (PP, ABS, PC) using the design of experiment approach. Similar trends such as the effects of a higher injection speed in decreasing the pressure work and of a lower temperature in decreasing pressure rate have been......Process monitoring of micro injection moulding (µIM) is of crusial importance to analyse the effect of different parameter settings on the process and to assess its quality. Quality factors related to cavity pressure can provide useful information directly connected with the dyanmics of the process...

  7. Two-phase coolant pump model of pressurized light water nuclear reactors

    International Nuclear Information System (INIS)

    Santos, G.A. dos; Freitas, R.L.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  8. Graphite beds for coolant filtration at high temperature

    International Nuclear Information System (INIS)

    Heathcock, R.E.; Lacy, C.S.

    1978-01-01

    High temperature filtration will be provided for new Ontario Hydro CANDU heat transport systems. Filtration has been shown to effectively reduce the concentration of circulating corrosion products in our heat transport systems, hence, minimizing the processes of activity transport. This paper will present one option we have for this application; Deep Bed Granular Graphite Filters. The filter system is described by discussing pertinent aspects of its development programme. The compatibility of the filter and the heat transport coolant are demonstrated by results from loop tests, both out- and in-reactor, and by subsequent results from a large filter installation in the NPD NGS heat transport system. (author)

  9. Interfacing systems loss of coolant accident (ISLOCA) pressure capacity methodology and Davis-Besse results

    International Nuclear Information System (INIS)

    Wesley, D.A.

    1991-01-01

    A loss of coolant accident resulting from the overpressurization by reactor coolant fluid of a system designed for low-pressure, low-temperature service has been identified as a potential contributor to nuclear power plant risk. In this paper, the methodology developed to assess the probability of failure as a function of internal pressure is presented, and sample results developed for the controlling failure modes and locations of four fluid systems at the Davis-Besse Plant are shown. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The variability in the probability of failure is included, and the estimated leak rates or leak areas are given for the controlling modes of failure. For this evaluation, all failures are based on quasistatic pressures since the probability of dynamic effects resulting from such causes as water hammer have been initially judged to be negligible for the Davis-Besse plant ISLOCA

  10. Heat transfer properties of organic coolants containing high boiling residues

    International Nuclear Information System (INIS)

    Debbage, A.G.; Driver, M.; Waller, P.R.

    1964-01-01

    Heat transfer measurements were made in forced convection with Santowax R, mixtures of Santowax R and pyrolytic high boiling residue, mixtures of Santowax R and CMRE Radiolytic high boiling residue, and OMRE coolant, in the range of Reynolds number 10 4 to 10 5 . The data was correlated with the equation Nu = 0.015 Re b 0.85 Pr b 0.4 with an r.m.s. error of ± 8.5%. The total maximum error arising from the experimental method and inherent errors in the physical property data has been estimated to be less than ± 8.5%. From the correlation and physical property data, the decrease in heat transfer coefficient with increasing high boiling residue concentration has been determined. It has been shown that subcooled boiling in organic coolants containing high boiling residues is a complex phenomenon and the advantages to be gained by operating a reactor in this region may be marginal. Gas bearing pumps used initially in these experiments were found to be unsuitable; a re-designed ball bearing system lubricated with a terphenyl mixture was found to operate successfully. (author)

  11. Analysis of loss-of-coolant accidents in pressurized water reactors

    International Nuclear Information System (INIS)

    Moldaschl, H.

    1982-01-01

    Analysis of loss-of-coolant accidents in pressurized water reactors -Quantification of the influence of leak size, control assembly worth, boron concentration and initial power by a dynamic operations criterion. Neutronic and thermohydraulic behaviour of a pressurized water reactor during a loss-of-coolant accident (LOCA) is mainly influenced by -change of fuel temperature, -void in the primary coolant. They cause a local stabilization of power density, that means that also in the case of small leaks local void is the main stabilization effect. As a consequence the increase of fuel temperature remains very small even under extremely hypothetical assumptions: small leak, positive reactivity feedback (positive coolant temperature coefficient, negative density coefficient) at the beginning of the accident and all control assemblies getting stuck. Restrictions which have been valid up to now for permitted start-up conditions to fulfill inherent safety requirements can be lossened substantially by a dynamic operations criterion. Burnable poisons for compensation of reactivity theorefore can be omitted. (orig.)

  12. Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

    1999-01-01

    During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

  13. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  14. Pressure Fluctuations in a Common-Rail Fuel Injection System

    Science.gov (United States)

    Rothrock, A M

    1931-01-01

    This report presents the results of an investigation to determine experimentally the instantaneous pressures at the discharge orifice of a common-rail fuel injection system in which the timing valve and cut-off valve were at some distance from the automatic fuel injection valve, and also to determine the methods by which the pressure fluctuations could be controlled. The results show that pressure wave phenomena occur between the high-pressure reservoir and the discharge orifice, but that these pressure waves can be controlled so as to be advantageous to the injection of the fuel. The results also give data applicable to the design of such an injection system for a high-speed compression-ignition engine.

  15. Reviving Abandoned Reservoirs with High-Pressure Air Injection: Application in a Fractured and Karsted Dolomite Reservoir

    Energy Technology Data Exchange (ETDEWEB)

    Robert Loucks; Stephen C. Ruppel; Dembla Dhiraj; Julia Gale; Jon Holder; Jeff Kane; Jon Olson; John A. Jackson; Katherine G. Jackson

    2006-09-30

    Despite declining production rates, existing reservoirs in the United States contain vast volumes of remaining oil that is not being effectively recovered. This oil resource constitutes a huge target for the development and application of modern, cost-effective technologies for producing oil. Chief among the barriers to the recovery of this oil are the high costs of designing and implementing conventional advanced recovery technologies in these mature, in many cases pressure-depleted, reservoirs. An additional, increasingly significant barrier is the lack of vital technical expertise necessary for the application of these technologies. This lack of expertise is especially notable among the small operators and independents that operate many of these mature, yet oil-rich, reservoirs. We addressed these barriers to more effective oil recovery by developing, testing, applying, and documenting an innovative technology that can be used by even the smallest operator to significantly increase the flow of oil from mature U.S. reservoirs. The Bureau of Economic Geology and Goldrus Producing Company assembled a multidisciplinary team of geoscientists and engineers to evaluate the applicability of high-pressure air injection (HPAI) in revitalizing a nearly abandoned carbonate reservoir in the Permian Basin of West Texas. The Permian Basin, the largest oil-bearing basin in North America, contains more than 70 billion barrels of remaining oil in place and is an ideal venue to validate this technology. We have demonstrated the potential of HPAI for oil-recovery improvement in preliminary laboratory tests and a reservoir pilot project. To more completely test the technology, this project emphasized detailed characterization of reservoir properties, which were integrated to access the effectiveness and economics of HPAI. The characterization phase of the project utilized geoscientists and petroleum engineers from the Bureau of Economic Geology and the Department of Petroleum

  16. AN EXPERIMENTAL NOX REDUCTION POTENTIAL INVESTIGATION OF THE PARTIAL HCCI APPLICATION, ON A HIGH PRESSURE FUEL INJECTION EQUIPPED DIESEL ENGINE BY IMPLEMENTING FUMIGATION OF GASOLINE PORT INJECTION

    OpenAIRE

    ERGENÇ, Alp Tekin; YÜKSEK, Levent; ÖZENER, Orkun; IŞIN, Övün

    2016-01-01

    This work investigates the effects of partial HCCI (Homogeneous charge compression ignition) application on today's modern diesel engine tail pipe NOx emissions. Gasoline fumigation is supplied via a port fuel injection system located in the intake port of DI(Direct injection) diesel engine to maintain partial HCCI conditions and also diesel fuel injected directly into the combustion chamber before TDC(Top dead center). A single cylinder direct injection diesel research engine equipped w...

  17. Reduced injection pressures using a compressed air injection technique (CAIT): an in vitro study.

    Science.gov (United States)

    Tsui, Ban C H; Knezevich, Mark P; Pillay, Jennifer J

    2008-01-01

    High injection pressures have been associated with intraneural injection and persistent neurological injury in animals. Our objective was to test whether a reported simple compressed air injection technique (CAIT) would limit the generation of injection pressures to below a suggested 1,034 mm Hg limit in an in vitro model. After ethics board approval, 30 consenting anesthesiologists injected saline into a semiclosed system. Injection pressures using 30 mL syringes connected to a 22 gauge needle and containing 20 mL of saline were measured for 60 seconds using: (1) a typical "syringe feel" method, and (2) CAIT, thereby drawing 10 mL of air above the saline and compressing this to 5 mL prior to and during injections. All anesthesiologists performed the syringe feel method before introduction and demonstration of CAIT. Using CAIT, no anesthesiologist generated pressures above 1,034 mm Hg, while 29 of 30 produced pressures above this limit at some time using the syringe feel method. The mean pressure using CAIT was lower (636 +/- 71 vs. 1378 +/- 194 mm Hg, P = .025), and the syringe feel method resulted in higher peak pressures (1,875 +/- 206 vs. 715 +/- 104 mm Hg, P = .000). This study demonstrated that CAIT can effectively keep injection pressures under 1,034 mm Hg in this in vitro model. Animal and clinical studies will be needed to determine whether CAIT will allow objective, real-time pressure monitoring. If high pressure injections are proven to contribute to nerve injury in humans, this technique may have the potential to improve the safety of peripheral nerve blocks.

  18. Image processing analysis of combustion for D. I. diesel engine with high pressure fuel injection. ; Effects of air swirl and injection pressure. Nensho shashin no gazo shori ni yoru koatsu funsha diesel kikan no nensho kaiseki. ; Swirl oyobi funsha atsuryoku no eikyo

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, I. (Japan Automobile Research Institute, Inc., Tsukuba (Japan)); Tsujimura, K.

    1994-02-25

    This paper reports an image processing analysis of combustion for a high-pressure direct injection diesel engine on the effects of air swirl and injection pressure upon combustion in the diesel engine. The paper summarizes a method to derive gas flow and turbulence strengths, and turbulent flow mixing velocity. The method derives these parameters by detecting movement of brightness unevenness on two flame photographs through utilizing the mutual correlative coefficients of image concentrations. Five types of combustion systems having different injection pressures, injection devices, and swirl ratios were used for the experiment. The result may be summarized as follows: variation in the average value of the turbulent flow mixing velocities due to difference in the swirl ratio is small in the initial phase of diffusion combustion; the difference is smaller in the case of high swirl ratio than in the case of low swirl ratio after the latter stage of the injection; the average value is larger with the higher the injection pressure during the initial stage of the combustion; after termination of the injection, the value is larger in the low pressure injection; and these trends agree with the trend in the time-based change in heat generation rates measured simultaneously. 6 refs., 14 figs., 2 tabs.

  19. Theoretical study of hydraulic jump during circular horizontal hot leg injection in pressurized water reactor

    International Nuclear Information System (INIS)

    El Hawary, Shehab; Abu-Elyazeed, Osayed S.M.; Fahmy, Adel Alyan; Meglaa, Khairy

    2016-01-01

    Highlights: • The model is developed to predict the occurrence of onset hydraulic jump in a circular pipe. • Theoretical results are in agreement with experimental results and theory. • Effects of diameter of the injection pipe, Froude number and injected coolant mass are studied. - Abstract: One important phenomenon occurring during Loss of Coolant Accident (LOCA) is Counter-Current Flow Limitation (CCFL). The incidence of such CCFL is introduced by the onset of hydraulic jump. In the present work, a one dimensional model was modified to fit circular hot channel. The model was used to study the factors affecting the initial Froude number, the location of the occurrence of the hydraulic jump, and the critical coolant flow depth during circular horizontal hot leg injection in US-APWR Mitsubishi Reactor. The results showed good agreement with published experimental data of the Upper Plenum Test Facility (UPTF) at Mannheim, Germany. It was found that higher injected coolant mass flow rate increases the initial Froude number, the location of the occurrence of the hydraulic jump, and the critical injection depth divided by the diameter of the injection pipe. Such behavior is thought to be due to the increase of the inertia force by increasing of the injected coolant mass flow rate and the inverse of the diameter of the injection pipe. It was found also that, the location of the occurrence of hydraulic jump increases with decreasing load effect. Therefore, these results reveal that the avoidance of CCFL as well as hydraulic jump through hot leg at maximum load can be achieved by decreasing the distance between the injection point and the pressure vessel to below 0.3 m, and with diameter of 4 in (10.16 cm) as the design diameter of the injection pipe in US-APWR Mitsubishi Reactor. Moreover, the maximum critical depth (56 cm) is less than the diameter of the hot leg (78.74 cm) at an injected coolant mass flow of 400 kg/s, and with diameter of 4 in (10.16 cm) as the

  20. Formation and hydraulic effects of deposits in high temperature sodium coolant systems

    International Nuclear Information System (INIS)

    Yunker, W.

    1976-01-01

    Deposition of sodium impurities in the high temperature (600 0 C), high flow (Reynolds Number approximately equal to 8 x 10 4 ) regions of a sodium coolant circuit is being studied to determine its possible hydraulic effects. Increases in flow impedance (pressure drop/volume flow 2 ) of up to 30 percent have been detected in an annular flow sensor. The apparatus and preliminary results of these tests are presented. Continuing tests are to specifically identify the materials involved and the system conditions under which the formations occur

  1. Analytical prediction on the pump-induced pulsating pressure in a reactor coolant pipe

    International Nuclear Information System (INIS)

    Lee, K.B.; Im, I.Y.; Lee, S.K.

    1992-01-01

    An analytical method is presented for predicting the amplitudes of pump-induced fluctuating pressures in a reactor coolant pipe using a linear transformation technique which reduces a homogeneous differential equation with non-homogeneous boundary conditions into a nonhomogeneous differential equation with homogeneous boundary conditions. At the end of the pipe, three types of boundary conditions are considered-open, closed and piston-spring supported. Numerical examples are given for a typical reactor. Comparisons of measured pressure amplitudes in the pipe with model prediction are shown to be in good agreement for the forcing frequencies. (author)

  2. A study on the effect of fluidic device installed in a safety injection tank on thermal-hydraulic phenomena of large break loss of coolant accident

    International Nuclear Information System (INIS)

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

    1999-03-01

    The performance of the Safety Injection Tank (SIT) with fluidic device (advanced SIT) is analyzed for the large break loss of coolant accident (LBLOCA) using RELAP5/MOD3.1-KREM. First the case is analyzed using the conventional SIT. Among various cases the case with 4-split downcomer, discharge coefficient Cd=0.6, MCP trip with reactor trip and break location of cold leg discharge side with the pressurizer is found to be the most limiting case. For the same condition, the advanced SIT results the similar PCT, however it can maintain adequately the liquid level in the downcomer. By changing the ECCS location from the current injection to the cold leg elevations, PCT is improved by 75 K. (Author). 6 refs., 4 tabs., 54 figs

  3. Investigation of a hydrogen mitigation system during large break loss-of-coolant accident for a two-loop pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dehjourian, Mehdi; Rahgoshay, Mohmmad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

  4. AMPTRACT: an algebraic model for computing pressure tube circumferential and steam temperature transients under stratified channel coolant conditions

    International Nuclear Information System (INIS)

    Gulshani, P.; So, C.B.

    1986-10-01

    In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution

  5. The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

    International Nuclear Information System (INIS)

    Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

    1993-01-01

    Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

  6. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  7. Performance Analysis of Multi Stage Safety Injection Tank

    International Nuclear Information System (INIS)

    Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo

    2015-01-01

    In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

  8. Performance Analysis of Multi Stage Safety Injection Tank

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

  9. Research on coolant radiochemistry

    International Nuclear Information System (INIS)

    Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

    2004-04-01

    The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

  10. Transmutation performance analysis on coolant options in a hybrid reactor system design for high level waste incineration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong-Hee; Siddique, Muhammad Tariq; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2015-11-15

    Highlights: • Waste transmutation performance was compared and analyzed for seven different coolant options. • Reactions of fission and capture showed big differences depending on coolant options. • Moderation effect significantly affects on energy multiplication, tritium breeding and waste transmutation. • Reduction of radio-toxicities of TRUs showed different trend to coolant choice from performance of waste transmutation. - Abstract: A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H{sub 2}O, D{sub 2}O, Na, PbBi, LiF-BeF{sub 2} and NaF-BeF{sub 2} applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are k{sub eff}, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H{sub 2}O and D{sub 2}O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF{sub 2} and NaF-BeF{sub 2} is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

  11. Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

    International Nuclear Information System (INIS)

    Boros, I.; Aszodi, A.

    2003-01-01

    The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

  12. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  13. Experimental study on spray break-up and atomization processes from GDI injector using high injection pressure up to 30 MPa

    International Nuclear Information System (INIS)

    Lee, Sanghoon; Park, Sungwook

    2014-01-01

    Highlights: • We obtain distribution of droplet velocity and diameter using PDPA system. • Transition of a jet break-up processes is visualized using Nd:Yag sheet laser system. • Elevated injection pressure can activate a jet break-up processes. • A limit in injection pressure to enhance droplet atomization is observed. -- Abstract: This paper focuses on the influence of injection pressures up to 30 MPa on single liquid jet break-up and atomization processes. For this purpose, a single jet from a multi-hole GDI injector has been characterized performing visualization and PDPA (phase Doppler particle analyzer) experiments. Using a thin sheet of light generated by a Nd:Yag laser and capturing a sequence of jet development images with a CCD camera, the internal structure was visualized. In order to quantify the droplet diameter and velocity, a 2-D PDPA system were carried out in addition to the spray visualization. Analyzing the images of the internal structure of jet and the result of PDPA, including droplet diameter and velocity distribution with increasing injection pressure up to 30 MPa, the elevated injection pressure on a jet break-up and atomization was characterized. Our experimental results show the existence of a leading edge of the jet observed at the initial stage of injection. This phenomenon revealed relatively large droplets ahead of the main jet then disappeared quickly as lose the droplets momentum. Furthermore, for all injection pressures, unique ‘branch-like structure’ was observed when the jet was fully developed. This structure had many counter rotating branches related to the effect of air-entrainment and rapidly broken down into droplet clusters and droplets. Especially, as increased injection pressure, the time to exhibit the structure and distance between two branches were decreased. In addition, based on the results of droplet diameter and velocity distribution at various injection pressures, we confirmed that the injection

  14. Analysis of an ultrasonic level device for in-core Pressurized Water Reactor coolant detection

    International Nuclear Information System (INIS)

    Johnson, K.R.

    1981-01-01

    A rigorous semi-empirical approach was undertaken to model the response of an ultrasonic level device (ULD) for application to in-core coolant detection in Pressurized Water Reactors (PWRs). An equation is derived for the torsional wave velocity v/sub t phi/ in the ULD. Existing data reduction techniques were analyzed and compared to results from use of the derived equation. Both methods yield liquid level measurements with errors of approx. 5%. A sensitivity study on probe performance at reactor conditions predicts reduced level responsivity from data at lower temperatures

  15. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  16. 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)

  17. Analysis of events related to cracks and leaks in the reactor coolant pressure boundary

    Energy Technology Data Exchange (ETDEWEB)

    Ballesteros, Antonio, E-mail: Antonio.Ballesteros-Avila@ec.europa.eu [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Sanda, Radian; Peinador, Miguel; Zerger, Benoit [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Negri, Patrice [IRSN: Institut de Radioprotection et de Sûreté Nucléaire (France); Wenke, Rainer [GRS: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

    2014-08-15

    Highlights: • The important role of Operating Experience Feedback is emphasised. • Events relating to cracks and leaks in the reactor coolant pressure boundary are analysed. • A methodology for event investigation is described. • Some illustrative results of the analysis of events for specific components are presented. - Abstract: The presence of cracks and leaks in the reactor coolant pressure boundary may jeopardise the safe operation of nuclear power plants. Analysis of cracks and leaks related events is an important task for the prevention of their recurrence, which should be performed in the context of activities on Operating Experience Feedback. In response to this concern, the EU Clearinghouse operated by the JRC-IET supports and develops technical and scientific work to disseminate the lessons learned from past operating experience. In particular, concerning cracks and leaks, the studies carried out in collaboration with IRSN and GRS have allowed to identify the most sensitive areas to degradation in the plant primary system and to elaborate recommendations for upgrading the maintenance, ageing management and inspection programmes. An overview of the methodology used in the analysis of cracks and leaks related events is presented in this paper, together with the relevant results obtained in the study.

  18. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  19. Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

    Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

  20. Mechanical energy yields and pressure volume and pressure time curves for whole core fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Coddington, P [United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1979-10-15

    In determining the damage consequences of a whole core Fuel-Coolant Interaction (FCI), one measure of the strength of a FCI that can be used and is independent of the system geometry is the constant volume mixing mechanical yield (often referred to as the Hicks-Menzies yield), which represents a near upper limit to the mechanical work of a FCI. This paper presents a recalculation of the Hicks-Menzies yields for UO{sub 2} and sodium for a range of initial fuel temperatures and fuel to coolant mass ratios, using recently published UO{sub 2} and sodium equation of state data. The work presented here takes a small number of postulated FCIs with as wide range as possible of thermal interaction parameters and determines their pressure-volume P(V) and pressure-time P(t) relations, using geometrical constraints representative of the reactor. Then by examining these P(V) and P(t) curves a representative pressure-relative volume curve or range of possible curves, for use in containment analysis, is recommended

  1. Coolant clean-up and recycle systems

    International Nuclear Information System (INIS)

    Ito, Takao.

    1979-01-01

    Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

  2. Post injection pressures in well treatments

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, G

    1967-06-05

    Behavior of wellhead pressure immediately after injection of liquids or slurries in well completion and workover treatments can often indicate the success of the operation. Since the rate of wellhead pressure build-down after injection is related to the permeability of the exposed formation to the treating fluid, interpretation of success or failure of the fluid to communicate with the reservoir is possible. Treatments designed to plug-up or clean-out formation flow channels can both be evaluated. Early appreciation can speed completion and workover operations. An explanation of the phenomena of increasing bottomhole treating pressure during fracture-type treatments, and the change in it throughout the life of a well, will result in better understanding of basic fracturing mechanics. On-the-job observations of decreasing rate of pressure build-down after increments of stage squeeze cementing will help the well-site engineer to vary the volume of increments of slurry and the duration of each stage.

  3. Behavior of pressure rise and condensation caused by water evaporation under vacuum at high temperature

    International Nuclear Information System (INIS)

    Takase, Kazuyuki; Kunugi, Tomoaki; Yamazaki, Seiichiro; Fujii, Sadao

    1998-01-01

    Pressure rise and condensation characteristics during the ingress-of-coolant event (ICE) in fusion reactors were investigated using the preliminary ICE apparatus with a vacuum vessel (VV), an additional tank (AT) and an isolation valve (IV). A surface of the AT was cooled by water at RT. The high temperature and pressure water was injected into the VV which was heated up to 250degC and pressure and temperature transients in the VV were measured. The pressure increased rapidly with an injection time of the water because of the water evaporation. After the IV was opened and the VV was connected with the AT, the pressure in the VV decreased suddenly. From a series of the experiments, it was confirmed that control factors on the pressure rise were the flushing evaporation and boiling heat transfer in the VV, and then, condensation of the vapor after was effective to the depressurization in the VV. (author)

  4. On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

    International Nuclear Information System (INIS)

    Melesed'Hospital, G.

    1983-01-01

    The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

  5. An improved apparatus for pressure-injecting fluid into trees

    Science.gov (United States)

    Garold F. Gregory; Thomas W. Jones

    1975-01-01

    Our original tree-injection apparatus was modified to be more convenient and efficient. The fluid reservoir consists of high-pressure plastic plumbing components. Quick couplers are used for all hose connections. Most important, the injector heads were modified for a faster and more convenient and secure attachment with double-headed nails.

  6. Effect of high-pressure homogenization preparation on mean globule size and large-diameter tail of oil-in-water injectable emulsions.

    Science.gov (United States)

    Peng, Jie; Dong, Wu-Jun; Li, Ling; Xu, Jia-Ming; Jin, Du-Jia; Xia, Xue-Jun; Liu, Yu-Ling

    2015-12-01

    The effect of different high pressure homogenization energy input parameters on mean diameter droplet size (MDS) and droplets with > 5 μm of lipid injectable emulsions were evaluated. All emulsions were prepared at different water bath temperatures or at different rotation speeds and rotor-stator system times, and using different homogenization pressures and numbers of high-pressure system recirculations. The MDS and polydispersity index (PI) value of the emulsions were determined using the dynamic light scattering (DLS) method, and large-diameter tail assessments were performed using the light-obscuration/single particle optical sensing (LO/SPOS) method. Using 1000 bar homogenization pressure and seven recirculations, the energy input parameters related to the rotor-stator system will not have an effect on the final particle size results. When rotor-stator system energy input parameters are fixed, homogenization pressure and recirculation will affect mean particle size and large diameter droplet. Particle size will decrease with increasing homogenization pressure from 400 bar to 1300 bar when homogenization recirculation is fixed; when the homogenization pressure is fixed at 1000 bar, the particle size of both MDS and percent of fat droplets exceeding 5 μm (PFAT 5 ) will decrease with increasing homogenization recirculations, MDS dropped to 173 nm after five cycles and maintained this level, volume-weighted PFAT 5 will drop to 0.038% after three cycles, so the "plateau" of MDS will come up later than that of PFAT 5 , and the optimal particle size is produced when both of them remained at plateau. Excess homogenization recirculation such as nine times under the 1000 bar may lead to PFAT 5 increase to 0.060% rather than a decrease; therefore, the high-pressure homogenization procedure is the key factor affecting the particle size distribution of emulsions. Varying storage conditions (4-25°C) also influenced particle size, especially the PFAT 5 . Copyright

  7. Experimental verification of integrated pressure suppression systems in fusion reactors at in-vessel loss-of-coolant events

    International Nuclear Information System (INIS)

    Takase, K.; Akimoto, H.

    2001-01-01

    An integrated ICE (Ingress-of-Coolant Event) test facility was constructed to demonstrate that the ITER safety design approach and design parameters for the ICE events are adequate. Major objectives of the integrated ICE test facility are: to estimate the performance of an integrated pressure suppression system; to obtain the validation data for safety analysis codes; and to clarify the effects of two-phase pressure drop at a divertor and the direct-contact condensation in a suppression tank. A scaling factor between the test facility and ITER-FEAT is around 1/1600. The integrated ICE test facility simulates the ITER pressure suppression system and mainly consists of a plasma chamber, vacuum vessel, simulated divertor, relief pipe and suppression tank. From the experimental results it was found quantitatively that the ITER pressure suppression system is very effective to reduce the pressurization due to the ICE event. Furthermore, it was confirmed that the analytical results of the TRAC-PF1 code can simulate the experimental results with high accuracy. (author)

  8. Aqueous Boric acid injection facility of PWR type reactor

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi; Iwami, Masao.

    1996-01-01

    If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

  9. On the transient pressure build-up in the full pressure safety shell of watercooled nuclear reactors after a loss of coolant accident

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1979-08-01

    The thermo-and fluid-dynamic processes in a multichamber full pressure safety containment during a loss of coolant accident have been investigated. Comparison of the calculations carried out with the computer programs, in which ZOCO VI was used as being representative of similar programs, with the experimental results pointed out discrepancies in the determination of time dependent pressure, pressure difference and temperature curves. This led to the development of a new theoretical model and a program COFLOW which pays particular attention to the fluid dynamic processes in the initial phase of a loss of coolant accident. It can also be used to determine the maximum containment pressure towards the end of a loss of coolant accident. Comparison of the COFLOW results with experiments has shown that COFLOW provides a model and a procedure by which the physical processes in a multichamber full pressure safety containment can be simulated satisfactorily

  10. Pressurization of a compartment due to the rupture of coolant piping

    International Nuclear Information System (INIS)

    Kot, C.A.; Hsieh, B.J.

    1993-01-01

    The pressurization and venting of enclosed compartments due to the accidental rupture of coolant piping is a safety problem common to many nuclear facilities. The processes associated with such an accident are very complex, involving, in general, transient multiphase flows, interactions and mixing between the incoming flows and the gases in the compartment, and heat transfer with the surroundings. Since pipe rupture is associated with many phenomenological uncertainties, elaborate 3-D thermal-hydraulic modeling and extensive calculational efforts are not warranted for many design applications. It is then more appropriate to rely. on simplified, global analysis approaches which can provide reasonably conservative estimates of the structural loads and flow processes, and which can readily be used in parameter/design studies. The objective of this paper is to present such an approach

  11. Fundamental study of a water jet injected into a vacuum vessel of fusion reactor under the ingress of coolant event

    International Nuclear Information System (INIS)

    Takase, Kazuyuki; Kunugi, Tomoaki; Seki, Yasushi; Kurihara, Ryouichi; Ueda, Shuzou

    1996-01-01

    As one of some transient sequences for the thermofluid safety in ITER, pressure rise and boiling heat transfer characteristics in a Tokamak vacuum vessel during an ingress of coolant event (ICE) are being investigated experimentally by using the preliminary ICE apparatus. The pressure rise rates in the vacuum vessel and the wall temperature distributions on the target plate were measured quantitatively and clarified at first. In addition, a two-phase flow under the ICE conditions was analyzed numerically for predicting the experimental results using one-dimensional transport equations and the drift-flux model. The experimental results were compared with the numerical results. It was found that the pressurization behavior during the ICE conditions could be estimated qualitatively by the present numerical analyses. 5 refs., 5 figs

  12. Development of in-situ laser cutting technique for removal of single selected coolant channel from pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Vishwakarma, S.C.; Upadhyaya, B.N.

    2016-01-01

    We report on the development of a pulsed Nd:YAG laser based cutting technique for removal of single coolant channel from pressurized heavy water reactor (PHWR). It includes development of special tools/manipulators and optimization of laser cutting process parameters for cutting of liner tube, end fitting, bellow lip weld joint, and pressure tube stubs. For each cutting operation, a special tool with precision motion control is utilized. These manipulators/tools hold and move the laser cutting nozzle in the required manner and are fixed on the same coolant channel, which has to be removed. This laser cutting technique has been successfully deployed for removal of selected coolant channels Q-16, Q-15 and N-6 of KAPS-2 reactor with minimum radiation dose consumption and in short time. (author)

  13. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-01-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

  14. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-12-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

  15. Research on Coolant Radiochemistry

    International Nuclear Information System (INIS)

    Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

    2007-06-01

    The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

  16. Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

  17. Consistent further development of the high pressure diesel fuel injection systems for passenger cars; Konsequente Weiterentwicklung der Hochdruck-Pkw-Dieseleinspritzsysteme

    Energy Technology Data Exchange (ETDEWEB)

    Warga, Johann; Pauer, Thomas; Boecking, Friedrich; Gerhardt, Juergen; Leonhard, Rolf [Robert Bosch GmbH, Stuttgart-Feuerbach (Germany). Diesel Systems

    2011-07-01

    Since the introduction of common rail technology in modern diesel engines for passenger cars there have been many changes and technological revolutions. Solely the continuous increase of the maximum injection pressure has remained unchanged as a guarantee for further engine performance improvement. Whether for down-sizing or for just simply increase the engine power or to reduce CO2 or to improve emissions: In all aspects the injection pressure can offer possible degrees of freedom. Besides, parallel to this continuous increase of injection pressure, the requirements concerning other injection system features have also continuously further developed. This paper focuses on the achievability of EU6 applications, among others, with the new Bosch 2000 bar solenoid valve injector, innovative nozzle technologies as e.g. with improved spray hole geometry or the modular concept common rail pump CP4. Current engine tests with pressures up to 2500 bar prove clearly the further advantages of pressure increase in diesel engines for passenger cars. In addition to the hydraulic components, system approaches in combination with electronic control, sensors and innovative control algorithms are increasingly in focus aiming to improve system accuracy and robustness. (orig.)

  18. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to

  19. System approach in the investigation of coolant parametrical oscillations in passive safety injection systems (PSIS)

    International Nuclear Information System (INIS)

    Proskouriakov, K.N.

    2001-01-01

    The use of thermal-hydraulic computer codes is an important part of the work programme for activities in the field of nuclear power plants (NPP) Safety Research as it will enable to define better the test configuration and parameter range extensions and to extrapolate the results of the small scale experiments towards full scale reactor applications. The CATHARE2, RELAP5, the WCOBRA/TRAC, and APROS codes are the estimate thermal hydraulic codes for the evaluation of large and small break loss of coolant accidents (LOCA). The relatively good agreement experimental data with the calculations have been presented. There was shown also some big mistakes in predicting distribution of flow when two phase are present. Model of parametrical oscillation (P.O.) worked out gives explanation for flow oscillations and indicates that the phenomenon of P.O. appears under certain combination of thermal-hydraulic parameters and structure of heat-removal system. (orig.)

  20. Transient performance analysis of pressurized safety injection tank with a partition

    International Nuclear Information System (INIS)

    Bae, Youngmin; Kim, Young In; Kim, Keung Koo

    2015-01-01

    Highlights: • Functional performance of safety injection tanks with a partition is evaluated. • Effects of key design parameters are scrutinized. • Distinctive features of the flow in multi-unit safety injection tanks are explored. - Abstract: A parametric study has been performed to evaluate the functional performance of a pressurized multi-unit safety injection tank, which would be considered as one of the candidates for a passive safety injection system in a nuclear power plant. The influences of key design parameters including the orifice size, initial gas fraction, and resistance coefficients and operating condition on the injection flow rate are scrutinized with a discussion of the relevant flow features such as the choked flow of gas through an orifice and two interconnected regions of differing gaseous pressure. The obtained results indicate that a multi-unit safety injection tank can passively control the injection flow rate and provide a stable safety injection over a relatively long period even in the case of drastic depressurization of a reactor coolant system

  1. Modelling the effect of injection pressure on heat release parameters and nitrogen oxides in direct injection diesel engines

    Directory of Open Access Journals (Sweden)

    Yüksek Levent

    2014-01-01

    Full Text Available Investigation and modelling the effect of injection pressure on heat release parameters and engine-out nitrogen oxides are the main aim of this study. A zero-dimensional and multi-zone cylinder model was developed for estimation of the effect of injection pressure rise on performance parameters of diesel engine. Double-Wiebe rate of heat release global model was used to describe fuel combustion. extended Zeldovich mechanism and partial equilibrium approach were used for modelling the formation of nitrogen oxides. Single cylinder, high pressure direct injection, electronically controlled, research engine bench was used for model calibration. 1000 and 1200 bars of fuel injection pressure were investigated while injection advance, injected fuel quantity and engine speed kept constant. The ignition delay of injected fuel reduced 0.4 crank angle with 1200 bars of injection pressure and similar effect observed in premixed combustion phase duration which reduced 0.2 crank angle. Rate of heat release of premixed combustion phase increased 1.75 % with 1200 bar injection pressure. Multi-zone cylinder model showed good agreement with experimental in-cylinder pressure data. Also it was seen that the NOx formation model greatly predicted the engine-out NOx emissions for both of the operation modes.

  2. High Blood Pressure

    Science.gov (United States)

    ... normal blood pressure 140/90 or higher is high blood pressure Between 120 and 139 for the top number, ... prehypertension. Prehypertension means you may end up with high blood pressure, unless you take steps to prevent it. High ...

  3. High Blood Pressure Facts

    Science.gov (United States)

    ... Stroke Heart Disease Cholesterol Salt Million Hearts® WISEWOMAN High Blood Pressure Facts Recommend on Facebook Tweet Share Compartir On ... Top of Page CDC Fact Sheets Related to High Blood Pressure High Blood Pressure Pulmonary Hypertension Heart Disease Signs ...

  4. High Blood Pressure (Hypertension)

    Science.gov (United States)

    ... Print Page Text Size: A A A Listen High Blood Pressure (Hypertension) Nearly 1 in 3 American adults has ... weight. How Will I Know if I Have High Blood Pressure? High blood pressure is a silent problem — you ...

  5. Reactor coolant cleanup device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

  6. Methodology for surge pressure evaluation in a water injection system

    Energy Technology Data Exchange (ETDEWEB)

    Meliande, Patricia; Nascimento, Elson A. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Dept. de Engenharia Civil; Mascarenhas, Flavio C.B. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Lab. de Hidraulica Computacional; Dandoulakis, Joao P. [SHELL of Brazil, Rio de Janeiro, RJ (Brazil)

    2009-07-01

    Predicting transient effects, known as surge pressures, is of high importance for offshore industry. It involves detailed computer modeling that attempts to simulate the complex interaction between flow line and fluid in order to ensure efficient system integrity. Platform process operators normally raise concerns whether the water injection system is adequately designed or not to be protected against possible surge pressures during sudden valve closure. This report aims to evaluate the surge pressures in Bijupira and Salema water injection systems due to valve closure, through a computer model simulation. Comparisons among the results from empirical formulations are discussed and supplementary analysis for Salema system were performed in order to define the maximum volumetric flow rate for which the design pressure was able to withstand. Maximum surge pressure values of 287.76 bar and 318.58 bar, obtained in Salema and Bijupira respectively, using empirical formulations have surpassed the operating pressure design, while the computer model results have pointed the greatest surge pressure value of 282 bar in Salema system. (author)

  7. Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

    International Nuclear Information System (INIS)

    Megahed, M.M.

    1992-01-01

    An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

  8. Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

    International Nuclear Information System (INIS)

    Elliot, A.J.; Stuart, C.R.

    2008-06-01

    An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

  9. Injection Pressure as a Marker of Intraneural Injection in Procedures of Peripheral Nerves Blockade

    Directory of Open Access Journals (Sweden)

    Ilvana Vučković

    2006-11-01

    Full Text Available The blockade of peripheral nerves carries a certain risk of unwanted complications, which can follow after unintentional intraneural injection of a local anesthetic. Up till today, the research of measuring injection pressure has been based on animal models, even though the histological structure of periphery nerve is different for animal and human population, so the application pressure which presages intraneural injection in human population is still unknown. As material in performing this study there have been used 12 Wistar rats and 12 delivered stillborns. After bilateral access to the median nerve, we applied 3 ml of 2% lidocaine with epinephrine, with the help of automatic syringe charger. The needle was at first placed perineural on one side, and then intraneural on the other side of both examination groups. During every application the pressure values were monitored using the manometer, and then they were analyzed by special software program BioBench. All perineural injections resulted with the pressure < or = 27.92 kPa, while the majority of intraneural injections were combined with the injectionpressure > or = 69.8 kPa. The difference between intraneural and perineural injection pressures for the two different examination groups (rats and delivered stillborns was not statistically significant (P>0.05. As prevention from intraneural injections today are in use two methods: the method of causing paresthesia or the method of using the peripheral nerve stimulator. However the nerve injury can still occur, independent from the technique used. If our results are used in clinical practice on human population, than the high injection pressure could be the markerof intraneural lodging of a needle.

  10. Coolant inlet device for nuclear reactors

    International Nuclear Information System (INIS)

    Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

    1969-01-01

    Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

  11. Numerical analysis of experiments with gas injection into liquid metal coolant

    International Nuclear Information System (INIS)

    Usov, E V; Lobanov, P D; Pribaturin, N A; Mosunova, N A; Chuhno, V I; Kutlimetov, A E

    2016-01-01

    Presented paper contains results of a numerical analysis of experiments with gas injection in water and liquid metal which have been performed at the Institute of Thermophysics Russian Academy of Science (IT RAS). Obtained experimental data are very important to predict processes that take place in the BREST-type reactor during the hypothetical accident with damage of the steam generator tubes, and may be used as a benchmark to validate thermo-hydraulic codes. Detailed description of models to simulate transport of gas phase in a vertical liquid column is presented in a current paper. Two-fluid model with closing relation for wall friction and interface friction coefficients was used to simulate processes which take place in a liquid during injection of gaseous phase. It has being shown that proposed models allow obtaining a good agreement between experimental data and calculation results. (paper)

  12. Hypertension (High Blood Pressure)

    Science.gov (United States)

    ... Safe Videos for Educators Search English Español Hypertension (High Blood Pressure) KidsHealth / For Teens / Hypertension (High Blood Pressure) What's ... rest temperature diet emotions posture medicines Why Is High Blood Pressure Bad? High blood pressure means a person's heart ...

  13. High pressure shaft seal

    International Nuclear Information System (INIS)

    Martinson, A.R.; Rogers, V.D.

    1980-01-01

    In relation to reactor primary coolant pumps, mechanical seal assembly for a pump shaft is disclosed which features a rotating seal ring mounting system which utilizes a rigid support ring loaded through narrow annular projections in combination with centering non-sealing O-rings which effectively isolate the rotating seal ring from temperature and pressure transients while securely positioning the ring to adjacent parts. A stationary seal ring mounting configuration allows the stationary seal ring freedom of motion to follow shaft axial movement up to 3/4 of an inch and shaft tilt about the pump axis without any change in the hydraulic or pressure loading on the stationary seal ring or its carrier. (author)

  14. Coolant leakage detecting device

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

    1995-01-01

    The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

  15. Residual heat removal pump and low pressure safety injection pump retrofit program

    International Nuclear Information System (INIS)

    Dudiak, J.G.; McKenna, J.M.

    1992-01-01

    Residual Heat Removal (RHR) and low pressure safety injection (LPSI) pumps installed in pressurized water-to-reactor power plants are used to provide low-head safety injection in the event of loss of coolant in the reactor coolant system. Because these pumps are subjected to rather severe temperature and pressure transients, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. Typically the pump impeller is mounted on an extended motor shaft (close-coupled configuration) and a mechanical seal is employed at the pump end of the shaft. Traditionally RHR and LPSI pumps have been a significant maintenance item for many utilities. Periodic mechanical seal of motor bearing replacement often is considered routine maintenance. The closed-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. This paper introduces a design modification developed to convert the close-coupled RHR and LPSI pumps to a coupled configuration

  16. Simulation of Spray Injection in the Pressurizer Using RELAP5

    Directory of Open Access Journals (Sweden)

    S. Dibyo

    2017-08-01

    Full Text Available A modeling research using Relap5 to assess the pressurizer of a pressurized water reactor(PWR power plant has been performed. The heater and water injection systems in the pressurizer system of the PWRare of greatimportance for system pressure control.The heater is designed to increase the pressure while the water sprayer injection is to perform depressurization. Most of studies conducted in the past mainly focused on determining the effects of nozzle spray design and droplet size using testing loops. The purpose of this simulation is to analyze the spray injection flow rate against the pressure characteristics of the pressurizer using RELAP5. Through this approach, the optimum injection flow rate of full scale plant pressurizer can be analyzed. The parameters investigated are pressure and temperature.In RELAP5, the pressurizer tank wasmodeled with six volume nodes and the heater was modeled by using heat structure. In the model, the sprayer takes water from the cold leg to inject it into the top of tank region.The resultsshowedthat the mass flow of about 4 kg/s is the mosteffectivevalueto limit pressure in the pressurizer to below 15.7 MPa. However, the flow rates of 8 kg/s and more cause overpressure. This simulation is usefulto complement the data related to the water flow rate injection systems of the pressurizer. Normal 0 false false false EN-US X-NONE X-NONE Temporal pore pressure induced stress changes during injection and depletion

    Science.gov (United States)

    Müller, Birgit; Heidbach, Oliver; Schilling, Frank; Fuchs, Karl; Röckel, Thomas

    2016-04-01

    Induced seismicity is observed during injection of fluids in oil, gas or geothermal wells as a rather immediate response close to the injection wells due to the often high-rate pressurization. It was recognized even earlier in connection with more moderate rate injection of fluid waste on a longer time frame but higher induced event magnitudes. Today, injection-related induced seismicity significantly increased the number of events with M>3 in the Mid U.S. However, induced seismicity is also observed during production of fluids and gas, even years after the onset of production. E.g. in the Groningen gas field production was required to be reduced due to the increase in felt and damaging seismicity after more than 50 years of exploitation of that field. Thus, injection and production induced seismicity can cause severe impact in terms of hazard but also on economic measures. In order to understand the different onset times of induced seismicity we built a generic model to quantify the role of poro-elasticity processes with special emphasis on the factors time, regional crustal stress conditions and fault parameters for three case studies (injection into a low permeable crystalline rock, hydrothermal circulation and production of fluids). With this approach we consider the spatial and temporal variation of reservoir stress paths, the "early" injection-related induced events during stimulation and the "late" production induced ones. Furthermore, in dependence of the undisturbed in situ stress field conditions the stress tensor can change significantly due to injection and long-term production with changes of the tectonic stress regime in which previously not critically stressed faults could turn to be optimally oriented for fault reactivation.

  17. Experimental investigation of coolant and poisoned moderator mixing due to a simulated pressure tube/calandria tube fishmouth rupturing an overpoisoned guaranteed shutdown state

    International Nuclear Information System (INIS)

    Mackinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

    1997-01-01

    During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The contents of the paper are as follows. First, the objectives of the experimental program are

  18. Effect of injection timing and injection pressure on the performance ...

    African Journals Online (AJOL)

    DR OKE

    This paper discusses the feasibility study on the utilization of biodiesel ester of Honge oil (EHO) in common rail direct injection. (CRDI) engine. Biodiesel of EHO has been obtained by transesterification process and characterization has been done. Existing single cylinder diesel engine fitted with conventional mechanical ...

  19. Numerical simulations on a high-temperature particle moving in coolant

    International Nuclear Information System (INIS)

    Li Xiaoyan; Shang Zhi; Xu Jijun

    2006-01-01

    This study considers the coupling effect between film boiling heat transfer and evaporation drag around a hot-particle in cold liquid. Taking momentum and energy equations of the vapor film into account, a transient single particle model under FCI conditions has been established. The numerical simulations on a high-temperature particle moving in coolant have been performed using Gear algorithm. Adaptive dynamic boundary method is adopted during simulating to matching the dynamic boundary that is caused by vapor film changing. Based on the method presented above, the transient process of high-temperature particles moving in coolant can be simulated. The experimental results prove the validity of the HPMC model. (authors)

  1. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2010-01-01

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  2. Vent clearing during a simulated loss-of-coolant accident in a Mark I boiling-water reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    In this test series, drywell pressurization rate, drywell overpressure, downcomer submergence, and overall vent system loss coefficient were varied to quantify the primary load sensitivities in the pressure suppression system. Extensive tests were conducted on a unique three-dimensional 1/5 scale model of the pressure suppression system a MARK-I BWR. They were focused on the initial or air cleaning phase of a hypothetical loss of coolant accident. As a result of the complete measurement system employed including multiple high speed cameras, the logical interrelationship between measured forces, measured pressures, and the hydrodynamic phenomena observed in high speed photographic pictures were established. The quantitative values from the 1/5 scale experiments can be applied to full scale plants using established scaling laws. (author)

  3. Validation of Computational Fluid Dynamics Calculation Using Rossendorf Coolant Mixing Model Flow Measurements in Primary Loop of Coolant in a Pressurized Water Reactor Model

    Directory of Open Access Journals (Sweden)

    Istvan Farkas

    2016-08-01

    Full Text Available The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively with experimental results.

  4. Method for removing cesium from aqueous liquid, method for purifying the reactor coolant in boiling water and pressurized water reactors and a mixed ion exchanged resin bed, useful in said purification

    International Nuclear Information System (INIS)

    Otte, J.N.A.; Liebmann, D.

    1989-01-01

    The invention relates to a method for removing cesium from an aqueous liquid, and to a resin bed containing a mixture of an anion exchange resin and cation exchange resin useful in said purification. In a preferred embodiment, the present invention is a method for purifying the reactor coolant of a presurized water or boiling water reactor. Said method, which is particularly advantageously employed in purifying the reactor coolant in the primary circuit of a pressurized reactor, comprises contacting at least a portion of the reactor coolant with a strong base anion exchange resin and the strong acid cation exchange resin derived from a highly cross-linked, macroporous copolymer of a monovinylidene aromatic and a cross-linking monomer copolymerizable therewith. Although the reactor coolant can sequentially be contacted with one resin type and thereafter with the second resin type, the contact is preferably conducted using a resin bed comprising a mixture of the cation and anion exchange resins. 1 fig., refs

  5. High blood pressure - children

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/007696.htm High blood pressure - children To use the sharing features on this page, please enable JavaScript. High blood pressure (hypertension) is an increase in the force of ...

  6. Preventing High Blood Pressure

    Science.gov (United States)

    ... Heart Disease Cholesterol Salt Million Hearts® WISEWOMAN Preventing High Blood Pressure: Healthy Living Habits Recommend on Facebook Tweet Share ... meal and snack options can help you avoid high blood pressure and its complications. Be sure to eat plenty ...

  7. High blood pressure - infants

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/007329.htm High blood pressure - infants To use the sharing features on this page, please enable JavaScript. High blood pressure (hypertension) is an increase in the force of ...

  8. High blood pressure medications

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/007484.htm High blood pressure medicines To use the sharing features on this page, please enable JavaScript. Treating high blood pressure will help prevent problems such as heart disease, ...

  9. High-pressure microbiology

    National Research Council Canada - National Science Library

    Michiels, Chris; Bartlett, Douglas Hoyt; Aertsen, Abram

    2008-01-01

    ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. High Hydrostatic Pressure Effects in the Biosphere: from Molecules to Microbiology * Filip Meersman and Karel Heremans . . . . . . . . . . . . 2. Effects...

  10. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  11. Improvement of lifetime availability through design, inspection, repair and replacement of coolant channels of Indian Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Rupani, B.B.; Sinha, R.K.

    1998-01-01

    This paper covers an overview of the work carried out for the life management of the coolant channels of Indian Pressurised Heavy Water Reactors. In order to improve maintainability of the coolant channels and reduce down time needed for periodical creep adjustment, improved design of channel hardware were developed. The modular insulation panel, designed as a substitute for the jig saw panels, reduces the time needed for accessing the space around the end-fitting significantly. A compact mechanical snubber has been developed to totally eliminate the need for periodic creep adjustment. In addition, the paper also describes the technologies developed for performing some special inspection, repair and replacement tasks for the coolant channels. These include systems for garter spring repositioning by Mechanical Flexing Technique for fresh reactors and Integrated Garter Spring Repositioning System for operating reactors. A tooling system, developed for in-situ retrieval of sliver scrape samples from pressure tubes, is also described. These samples can be analysed in laboratories to yield valuable information on hydrogen concentration in pressure tube material. The current and planned activities towards development of technologies for improvement of the life time availability of the power plants are addressed. (author)

  12. Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kanzleiter, T.

    1975-10-01

    The phenomena occuring within a containment during a LOCA are currently investigated through experiments with a modelcontainment at Battelle-Institut Frankfurt on behalf of the Bundesministerium fuer Forschung und Technologie, Bonn. The experimental results are to be compared with the results of model calculations in order to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model-containment. The model-containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross section. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiment a PWR-configuration with nine compartments has been istalled. The model scale of the compartment volumes and the overflow areas are about 1:64 compared to the 1,200-MW-PWR-plant Biblis A. Later investigations will also include BWR-experiments and experiments leading to an extremely high load on special containment structures. (orig.) [de

  13. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

  14. One-phase and two-phase homologous curves for coolant pumps of the pressurized light water nuclear reactors

    International Nuclear Information System (INIS)

    Santos, G.A. dos.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The single-phase pump characteristics are an essential feature for operational transients studies, for example, the shut-down and start-up of pump. These parameters, in terms of the homologous curves, set up the complete performance of the pump and are input for transients and accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the single-phase and two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  15. Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

    International Nuclear Information System (INIS)

    Lydell, B.

    1999-05-01

    This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research

  16. Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B. [RSA Technologies, Vista, CA (United States)

    1999-05-01

    This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research 41 refs, figs, tabs

  17. Correlation of analysis with high level vibration test results for primary coolant piping

    International Nuclear Information System (INIS)

    Park, Y.J.; Hofmayer, C.H.; Costello, J.F.

    1992-01-01

    Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was gradually increased up to the limit of the shaking table and significant plastic strains, as well as cracking, were induced in the piping. To fully utilize the test results, NRC/BNL sponsored a project to develop corresponding analytical predictions for the nonlinear dynamic response of the piping for selected test runs. The analyses were performed using both simplified and detailed approaches. The simplified approaches utilize a linear solution and an approximate formulation for nonlinear dynamic effects such as the use of a deamplification factor. The detailed analyses were performed using available nonlinear finite element computer codes, including the MARC, ABAQUS, ADINA and WECAN codes. A comparison of various analysis techniques with the test results shows a higher prediction error in the detailed strain values in the overall response values. A summary of the correlation analyses was presented before the BNL. This paper presents a detailed description of the various analysis results and additional comparisons with test results

  18. High-pressure apparatus

    NARCIS (Netherlands)

    Schepdael, van L.J.M.; Bartels, P.V.; Berg, van den R.W.

    1999-01-01

    The invention relates to a high-pressure device (1) having a cylindrical high-pressure vessel (3) and prestressing means in order to exert an axial pressure on the vessel. The vessel (3) can have been formed from a number of layers of composite material, such as glass, carbon or aramide fibers which

  19. Influence of fuel injection pressures on Calophyllum inophyllum methyl ester fuelled direct injection diesel engine

    International Nuclear Information System (INIS)

    Nanthagopal, K.; Ashok, B.; Karuppa Raj, R. Thundil

    2016-01-01

    Highlights: • Effect of injection pressure of Calophyllum inophyllum biodiesel is investigated. • Engine characteristics of 100% Calophyllum inophyllum biodiesel has been performed. • Calophyllum inophyllum is a non-edible source for biodiesel production. • Increase in injection pressure of biodiesel, improves the fuel economy. • Incylinder pressure characteristics of biodiesel follows similar trend as of diesel. - Abstract: The trend of using biodiesels in compression ignition engines have been the focus in recent decades due to the promising environmental factors and depletion of fossil fuel reserves. This work presents the effect of Calophyllum inophyllum methyl ester on diesel engine performance, emission and combustion characteristics at different injection pressures. Experimental investigations with varying injection pressures of 200 bar, 220 bar and 240 bar have been carried out to analyse the parameters like brake thermal efficiency, specific fuel consumption, heat release rate and engine emissions of direct injection diesel engine fuelled with 100% biodiesel and compared with neat diesel. The experimental results revealed that brake specific fuel consumption of C. inophyllum methyl ester fuelled engine has been reduced to a great extent with higher injection pressure. Significant reduction in emissions of unburnt hydrocarbons, carbon monoxide and smoke opacity have been observed during fuel injection of biodiesel at 220 bar compared to other fuel injection pressures. However oxides of nitrogen increased with increase in injection pressures of C. inophyllum methyl ester and are always higher than that of neat diesel. In addition the combustion characteristics of biodiesel at all injection pressures followed a similar trend to that of conventional diesel.

  20. Qualifying Elbow Meters for High Pressure Flow Measurements in an Operating Nuclear Power Plant

    International Nuclear Information System (INIS)

    Chan, A.M.; Maynard, K.J.; Ramundi, J.; Wiklung, E.

    2006-01-01

    To support the installation and use of elbow meters to measure the high pressure emergency coolant injection flow in an operating nuclear station, a test program was performed to qualify: (i) the 'hot' tapping procedure for field application and (ii) the use of elbow meters for accurate flow measurements over the full range of station ECI flow conditions. This paper describes the design conditions and major components of a flow loop used for the elbow meter calibrations. Typical test results are presented and discussed. (authors)

  1. High-temperature process heat reactor with solid coolant and radiant heat exchange

    International Nuclear Information System (INIS)

    Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

    1984-01-01

    The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion

  2. High blood pressure - adults

    Science.gov (United States)

    ... pressure is found. This is called essential hypertension. High blood pressure that is caused by another medical condition or medicine you are taking is called secondary hypertension. Secondary hypertension may be due to: Chronic ...

  3. Optimization of injection pressure for a compression ignition engine ...

    African Journals Online (AJOL)

    user

    injection and atomization and contributes to incomplete combustion, nozzle clogging, ... this non edible oil may be an appropriate substitute for diesel fuel. ... The effect of injector opening pressure on the performance of a jatropha oil fuelled ...

  4. Study of air entrainment in high pressure spray: optics diagnostics and application to the Diesel injection; Etude de l'entrainement d'air dans un spray haute pression: diagnostics optiques et application a l'injection diesel

    Energy Technology Data Exchange (ETDEWEB)

    Arbeau, A.

    2004-12-15

    The actual development of the engine must reply to a will of fuel consumption reduction and to norms more and more strict concerning the pollutant emissions. Although the Diesel engines are efficient, the NO{sub x} and particle emissions mainly come from the existence of wealthy fuel zone preventing an optimal combustion. Therefore, the air / fuel mixing preparation, highly controlled by the air entrainment in spray, is essential. In this context, we have developed metrological tools in order to analyse the air entrainment mechanism in a dense spray. The Particle Image Velocimetry (PIV) technique is first applied to a conical spray with an injection pressure less than 100 bars to study the air entrainment in spray. A transfer of the methodologies allows then the characterisation and the understanding of the air entrainment mechanism in high pressure full spray (injection pressure less than 1600 bars) type Diesel one. The influence of injection parameters (injection pressure and back pressure) on the mixing rate is studied. The increase of the injection pressure from 800 to 1600 bars implies an increase of the mixing rate of 60 %. Moreover, the thermodynamic conditions of the ambient air, simulated by the chamber back pressure, widely favours the mixing rate. Actually, this latter increases of 350 % when the chamber back pressure varies from 1 to 7 bars. The experimental results do not follow classical laws of air entrainment in one-phase flow jet with variable density, but are in good agreement with an integral model for air entrainment in an axisymmetric full spray. Finally, the Fluorescence Particle Image Velocimetry (FPIV) is introduced in order to extend the PIV application field in dense two-phase flows. (author)

  5. Research on loss of coolant accident of pressurized-water reactor based on PSO algorithm

    International Nuclear Information System (INIS)

    Ma Jie; Guo Lifeng; Peng Qiao

    2012-01-01

    In order to improve the diagnosis performance of Loss of Coolant Accident (LOCA), based on Back Propagation (BP) algorithm study, a fault diagnosis network is established based on Particle Swarm Optimization (PSO) algorithm in this paper. The PSO algorithm is used to train the weights and the thresholds of neural network, which can conquer part convergence problem of BP algorithm. The test results show that the diagnosis network has higher accuracy of LOCA. (authors)

  6. Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Carl, H.; Kubis, K.

    1979-01-01

    Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

  7. CONTEMPT: computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1978-04-01

    The CONTEMPT code is used by Babcock and Wilcox for containment analysis following a postulated loss of coolant accident. An additional model is described which is used for the calculation of long term post reflood mass and energy releases to the containment that is used for the containment design basis LOCA calculations. These calculations maximize the rate of energy flow to the containment. The mass and energy data are given to the containment designer for use in calculating the containment building design pressure and temperature and in sizing containment heat removal equipment

  8. Linear titration plot for the determination of boron in the primary coolant of a pressurized water reactor

    International Nuclear Information System (INIS)

    Midgley, D.; Gatford, C.

    1992-01-01

    A linear titration plot method has been devised for the determination of boron as boric acid in partly neutralized solution, such as occurs in the primary coolant of pressurized water reactors. The total boron and the alkali in the sample are determined simultaneously. Although it is not essential to add mannitol in this method, it is more accurate when the solution is saturated with mannitol. Comparisons are made with other modes of titration: Gran plots, first and second differential potentiometric titrations and indicator titrations. None of these gives the total boron directly in partly neutralized solutions. (author)

  9. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  10. Unique rod lens/video system designed to observe flow conditions in emergency core coolant loops of pressurized water reactors

    International Nuclear Information System (INIS)

    Carter, G.W.

    1979-01-01

    Techniques and equipment are described which are used for video recordings of the single- and two-phase fluid flow tests conducted with the PKL Spool Piece Measurement System designed by Lawrence Livermore Laboratory and EG and G Inc. The instrumented spool piece provides valuable information on what would happen in pressurized water reactor emergency coolant loops should an accident or rupture result in loss of fluid. The complete closed-circuit television video system, including rod lens, light supply, and associated spool mounting fixtures, is discussed in detail. Photographic examples of test flows taken during actual spool piece system operation are shown

  11. Analysis of the effects of the pressure wave generated in loss of coolant accidents in reactor vessels

    International Nuclear Information System (INIS)

    Valero Martinez, M.

    1980-01-01

    The increasing demands in the field of ''Nuclear Safety'', obliges to a perfect knowledge of the causes and effects of every possible accident in a nuclear power plant. In this paper will be analysed the effects of the pressure wave appearing in a LOCA (Loss of collant accident). The pressure wave could deform the following structures: core barrel wall, cover and bottom, control rods and safety coolant system. Any change of the geometry of these structures could provoke and incorrect system reaction after the accident has happened. The basis and hypothesis for the theoretical analysis will be exposed. The structures are considered to be rigid. A typical boiling water be analysed and the developed theory will be verified in comparations with experimental results and the results obtained with some others models. Due to the easy application and short calculation time of the created programmes, they are recommended for parametrical calculations in the analysis of the pressurized water reactors and boiling water reactors. (author)

  12. High Blood Pressure (Hypertension)

    Science.gov (United States)

    ... other risk factors, like diabetes, you may need treatment. How does high blood pressure affect pregnant women? A few women will get ... HIV, Birth Control Heart Health for Women Pregnancy Menopause More Women's Health ... High Blood Pressure--Medicines to Help You Women and Diabetes Heart ...

  13. High-pressure crystallography

    Science.gov (United States)

    Katrusiak, A.

    2008-01-01

    The history and development of high-pressure crystallography are briefly described and examples of structural transformations in compressed compounds are given. The review is focused on the diamond-anvil cell, celebrating its 50th anniversary this year, the principles of its operation and the impact it has had on high-pressure X-ray diffraction.

  14. Research on natural gas fuel injection system. Development of high-performance pressure regulator; Tennen gas yo nenryo funsha system no kenkyu kaihatsu. 1. Tennen gas nenryo funshayo no koseino regulator kaihatsu

    Energy Technology Data Exchange (ETDEWEB)

    Kato, S; Ishii, M; Takigawa, B; Makabe, K; Harada, S; Ono, H [Nippon Carburetor Co. Ltd., Tokyo (Japan)

    1997-10-01

    With the aim of further reducing the exhaust emissions of natural-gas vehicles, vigorous research and development work is under way today on multi point gas injection (MPGI) system. In this studies, a high-performance pressure regulator, which is one of the main components of this MPGI system, has been newly developed. The results showed that a significantly better accuracy of the regulated pressure level using this regulator was obtained under the wide range of operating conditions, including instantaneously greater changes of fuel flow rate. In addition, the advanced studies of gaseous fuel injectors (GFIs) would be also conducted. 4 refs., 8 figs.

  15. Synthesis of ethylene glycol-treated Graphene Nanoplatelets with one-pot, microwave-assisted functionalization for use as a high performance engine coolant

    International Nuclear Information System (INIS)

    Amiri, Ahmad; Sadri, Rad; Shanbedi, Mehdi; Ahmadi, Goodarz; Kazi, S.N.; Chew, B.T.; Zubir, Mohd Nashrul Mohd

    2015-01-01

    Highlights: • A potentially mass production method is introduced for preparing EG-treated GNP. • A promising car radiator coolant in the presence of neutral media synthesized. • Car engine can work in lower temperature via high-performance coolant. • The ratio of convective to conductive heat transfer is unique. • New economical product with high performance index is introduced. - Abstract: An electrophilic addition reaction under microwave irradiation was developed as a promising, quick and cost-effective approach to functionalize Graphene Nanoplatelets (GNP) with ethylene glycol (EG). EG-treated GNP was synthesized to reach a promising dispersibility in the water–EG media without negative effects of acid-treatment. Surface functionality groups and the morphology of chemically-functionalized GNP were characterized by the vibration spectroscopies, temperature-programmed study, and microscopic method. Despite the fact that the main structures of GNP were remained reasonably intact, characterization results consistently verified the functionalization of GNP with EG functionalities. As new kinds of high-performance engine coolant, the EG-treated GNP based water–EG coolant (GNP-WEG) was prepared and its thermo-physical and rheological properties are evaluated. In particular, the thermal conductivity, viscosity, specific heat capacity, and density of all samples were experimentally measured to evaluate the thermal performance of the GNP-WEG coolant. The data showed insignificant increases in the pressure drop at different temperatures and concentrations, low friction factor, lack of corrosive condition, and the performance index larger than 1. In addition, no momentous change in the pumping power in the presence of GNP-WEG confirmed that it can be an appropriate alternative coolant for different thermal equipment in terms of economy and performance

  16. Influence of Zn injection on corrosion behavior and oxide film characteristics of 304 stainless steel in borated and lithiated high temperature water

    International Nuclear Information System (INIS)

    Wu, Xinqiang; Liu, Xiahe; Han, En-Hou; Ke, Wei

    2012-09-01

    Water chemistry of the reactor coolant system plays a major role in maintaining safety and reliability of light water reactor nuclear power plants (NPPs). Zn water chemistry into pressurized water reactors (PWRs) in order to reduce the radiation buildup in primary coolant system has been widely applied, and the reduction effect has been experimentally confirmed. Zn injection can also lessen the corrosion phenomena in high temperature pressurized water by changing oxide films formed on components materials. Both the radiation buildup and material corrosion resistance in PWR coolant system are closely dependent on the oxide films formed. However, the influence of Zn injection on the chemical composition and structure of the oxide films on their protective properties is still a matter of considerable debate. The influence of Zn injection on corrosion inhibition and environmental degradation has not been fully clarified yet. Therefore, the understanding of corrosion behaviour, oxide film characteristics and their protective property is of significance to clarify the environmentally assisted material failure problems in NPPs. In the present work, oxide films formed on nuclear-grade 304 SS exposed to borated and lithiated high temperature water environments at 300 deg. C up to 4000 h with or without 10 ppb Zn injection were investigated ex-situ. Without Zn injection, the oxide films mainly consisted of Fe 3 O 4 and FeCr 2 O 4 . With Zn injection, ZnFe 2 O 4 and ZnCr 2 O 4 were detected in the oxide films at the initial stage of immersion and ZnCr 2 O 4 became dominant after long-term immersion. It was believed that the above Zn-Fe and Zn-Cr spinel oxides were formed by substitution reactions between Zn 2+ and Fe 2+ . At the initial stage of immersion, water chemistry significantly affected the formation of the oxide films. Once a stable oxide film formed, it is rather difficult to change its structure through changing water chemistry. The potential-pH diagrams for Zn

  17. Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

  18. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    Science.gov (United States)

    Xie, Huaqing; Li, Yang; Yu, Wei

    2010-05-01

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al 2O 3, ZnO, TiO 2, and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al 2O 3, and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  19. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    International Nuclear Information System (INIS)

    Xie Huaqing; Li Yang; Yu Wei

    2010-01-01

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al 2 O 3 , ZnO, TiO 2 , and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al 2 O 3 , and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  20. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    Energy Technology Data Exchange (ETDEWEB)

    Xie Huaqing, E-mail: hqxie@eed.sspu.c [School of Urban Development and Environmental Engineering, Shanghai Second Polytechnic University, Shanghai 201209 (China); Li Yang; Yu Wei [School of Urban Development and Environmental Engineering, Shanghai Second Polytechnic University, Shanghai 201209 (China)

    2010-05-31

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al{sub 2}O{sub 3}, ZnO, TiO{sub 2}, and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al{sub 2}O{sub 3}, and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  1. Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kanzleiter, T.F.

    1976-01-01

    For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1 : 64 compared to the 1200 MW PWR plant Biblis A. (Auth.)

  2. Chaotic behavior of water column oscillator simulating pressure balanced injection system in passive safety reactor

    International Nuclear Information System (INIS)

    Morimoto, Y.; Madarame, H.; Okamoto, K.

    2001-01-01

    Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor called the System-integrated Pressurized Water Reactor (SPWR). In a loss of coolant accident, the Pressurizing Line (PL) and the Injection Line (IL) are passively opened. Vapor generated by residual heat pushes down the water level in the Reactor Vessel (RV). When the level is lower than the inlet of the PL, the vapor is ejected into the Containment Vessel (CV) through the PL. Then boronized water in the CV is injected into the RV through the IL by the static head. In an experiment using a simple apparatus, gas ejection and water injection were found to occur alternately under certain conditions. The gas ejection interval was observed to fluctuate considerably. Though stochastic noise affected the interval, the experimental results suggested that the large fluctuation was produced by an inherent character in the system. A set of piecewise linear differential equations was derived to describe the experimental result. The large fluctuation was reproduced in the analytical solution. Thus it was shown to occur even in a deterministic system without any source of stochastic noise. Though the derived equations simulated the experiment well, they had ten independent parameters governing the behavior of the solution. There appeared chaotic features and bifurcation, but the analytical model was too complicated to examine the features and mechanism of bifurcation. In this study, a new simple model is proposed which consists of a set of piecewise linear ordinary differential equations with only four independent parameters. (authors)

  3. Pressure changes in the plasma sheet during substorm injections

    International Nuclear Information System (INIS)

    Kistler, L.M.; Moebuis, E.; Baumjohann, W.; Paschmann, G.; Hamilton, D.C.

    1992-01-01

    The authors have determined the particle pressure and total pressure as a function of radial distance in the plasma sheet for periods before and after the onset of substorm-associated ion enhancements over the radial range 7-19 R E . They have chosen events occurring during times of increasing magnetospheric activity, as determined by an increasing AE index, in which a sudden increase, or injection, of energetic particle flux is observed. During these events the particle energy of maximum contribution to the pressure increases from about 12 to about 27 keV. In addition, the particle pressure increases, and the magnetic pressure decreases, with the total pressure only changing slightly. For radial distances of less than 10 R E the total pressure tends to increase with the injection, while outside 10 R E it tends to decrease or remain the same. Because the fraction of the pressure due to particles has increased and higher energies are contributing to the pressure, a radial gradient is evident in the postinjection, but not preinjection, flux measurements. These observations show that the simulations appearance of energetic particles and changes in the magnetic field results naturally from pressure balance and does not necessarily indicate that the local changing field is accelerating the particles. The changes in the total pressure outside 10 R E are consistent with previous measurements of pressure changes at substorm onset and can be understood in terms of the unloading of energy in the magnetotail and the resulting change in the magnetic field configuration

  4. High temperature and high performance light water cooled reactors operating at supercritical pressure, research and development

    International Nuclear Information System (INIS)

    Oka, Y.; Koshizuka, S.; Katsumura, Y.; Yamada, K.; Shiga, S.; Moriya, K.; Yoshida, S.; Takahashi, H.

    2003-01-01

    The concept of supercritical-pressure, once-through coolant cycle nuclear power plant (SCR) was developed at the University of Tokyo. The research and development (R and D) started worldwide. This paper summarized the conceptual design and R and D in Japan. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical fossil fired power plants (FPP) in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. Boiling evolved from circular boilers, water tube boilers and once-through boilers. It is the reactor version of the once-through boiler. The development from LWR to SCR follows the history of boilers. The goal of the R and D should be the capital cost reduction that cannot be achieved by the improvement of LWR. The reactor can be used for hydrogen production either by catalysis and chemical decomposition of low quality hydrocarbons in supercritical water. The reactor is compatible with tight lattice fast core for breeders due to low outlet coolant density, small coolant flow rate and high head coolant pumps

  5. Low pressure powder injection moulding of stainless steel powders

    Energy Technology Data Exchange (ETDEWEB)

    Zampieron, J.V.; Soares, J.P.; Mathias, F.; Rossi, J.L. [Powder Processing Center CCP, Inst. de Pesquisas Energeticas e Nucleares, Sao Paulo, SP (Brazil); Filho, F.A. [IPEN, Inst. de Pesquisas Energeticas e Nucleares, Cidade Univ., Sao Paulo, SP (Brazil)

    2001-07-01

    Low-pressure powder injection moulding was used to obtain AISI 316L stainless steel parts. A rheological study was undertaken using gas-atomised powders and binders. The binders used were based on carnauba wax, paraffin, low density polyethylene and microcrystalline wax. The metal powders were characterised in terms of morphology, particle size distribution and specific surface area. These results were correlated to the rheological behaviour. The mixture was injected in the shape of square bar specimens to evaluate the performance of the injection process in the green state, and after sintering. The parameters such as injection pressure, viscosity and temperature were analysed for process optimisation. The binders were thermally removed in low vacuum with the assistance of alumina powders. Debinding and sintering were performed in a single step. This procedure shortened considerably the debinding and sintering time. (orig.)

  6. Investigations of the reflood-phase after a loss-of-coolant-accident of an advanced pressurized water reactor (APWR)

    International Nuclear Information System (INIS)

    Schumann, S.; Oldekop, W.

    1983-01-01

    Differences between a high converting advanced pressurized-water reactor (APWR) and a conventional PWR, which are relevant to the reflood-phase after LOCA are presented. The used code and its verification by PWR-reflood experiments is explained. Comparative calculations for APWR and PWR with several conservative assumptions for example cold-leg-injection only, yield nearly the same maximum midplane-temperatures for the average-channel. For the APWR, however, the upper half of the rod shows higher temperatures. Quenchfront and core-water-level increase more slowly. The differences in the reflood-thermohydraulics are analysed in detail. A conservative hot-channel calculation shows maximum temperatures of about 920 0 C. Finally the influence of conservative assumptions is described and the necessity of experiments pointed out. (orig.)

  7. High Blood Pressure

    Science.gov (United States)

    ... kidney disease, diabetes, or metabolic syndrome Read less Unhealthy lifestyle habits Unhealthy lifestyle habits can increase the risk of high blood pressure. These habits include: Unhealthy eating patterns, such as eating too much sodium ...

  8. High Blood Pressure

    Science.gov (United States)

    ... factors Diabetes High blood pressure Family history Obesity Race/ethnicity Full list of causes and risk factors ... give Give monthly Memorials and tributes Donate a car Donate gently used items Stock donation Workplace giving ...

  9. High-pressure tritium

    International Nuclear Information System (INIS)

    Coffin, D.O.

    1976-01-01

    Some solutions to problems of compressing and containing tritium gas to 200 MPa at 700 0 K are discussed. The principal emphasis is on commercial compressors and high-pressure equipment that can be easily modified by the researcher for safe use with tritium. Experience with metal bellows and diaphragm compressors has been favorable. Selection of materials, fittings, and gauges for high-pressure tritium work is also reviewed briefly

  10. Optimum injection pressure of a cavitating jet on introduction of compressive residual stress into stainless steel

    International Nuclear Information System (INIS)

    Soyama, Hitoshi; Nagasaka, Kazuya; Takakuwa, Osamu; Naito, Akima

    2011-01-01

    In order to mitigate stress corrosion cracking of components used for nuclear power plants, introduction of compressive residual stress into sub-surface of the components is an effective maintenance method. The introduction of compressive residual stress using cavitation impact generated by injecting a high speed water jet into water was proposed. Water jet peening is now applying to reduce stress corrosion cracking of shrouds in the nuclear power plants. However, accidental troubles such as dropping off the components and cutting of the pipes by the jet occurred at the maintenance. In order to peen by the jet without damage, optimum injection pressure of the jet should be revealed. In the case of 'cavitation peening', cavitation is generated by injecting the high speed water jet into water. As working pressure at the cavitation peening is the pressure at cavitation bubble collapse, the injection pressure of the jet is not main parameter. The cavitation impact is increasing with the scale of the jet, i.e., scaling effect of the cavitation. It was revealed that the large scale jet at low injection pressure can introduce compressive residual stress into stainless steel comparing with the small scale jet at high injection pressure. As expected, a water jet at high injection pressure might make damage of the components. Namely, in order to avoid damage of the components, the jet at the low injection pressure will be suit for the introduction of compressive residual stress. In the present paper, in order to make clear optimum injection pressure of the cavitating jet for the introduction of compressive residual stress without damage, the residual stress of stainless steel treated by the jet at various injection pressure was measured by using an X-ray diffraction method. The injection pressure of the jet p 1 was varied from 5 MPa to 300 MPa. The diameter of the nozzle throat of the jet d was varied from 0.35 mm to 2.0 mm. The residual stress changing with depth was

  11. Turbulent heat transfer in a coolant channel of a pressurized water reactor (PWR) core

    International Nuclear Information System (INIS)

    Kumar, Sanjeev; Saha, Arun K.; Munshi, Prabhat

    2016-01-01

    Exact predictions in nuclear reactors are more crucial, because of the safety aspects. It necessitates the appropriate modeling of heat transfer phenomena in the reactors core. A two-dimensional thermal-hydraulics model is used to study the detailed analysis of the coolant region of a fuel pin. Governing equations are solved using Marker and Cell (MAC) method. Standard wall functions k-ε turbulence model is incorporated to consider the turbulent behaviour of the flow field. Validation of the code and a few results for a typical PWR running at normal operating conditions reported earlier. There were some discrepancies in the old calculations. These discrepancies have been resolved and updated results are presented in this work. 2D thermal-hydraulics model results have been compared with the 1D thermal-hydraulics model results and conclusions have been drawn. (author)

  12. Frontier between medium and large break loss of coolant accidents of pressurized water reactor

    Science.gov (United States)

    Kim, Taewan

    2017-10-01

    In order to provide the probabilistic safety assessment with more realistic condition to calculate the frequency of the initiating event, a study on the frontier between medium-break and large-break loss-of-coolant-accidents has been performed by using best-estimate thermal hydraulic code, TRACE. A methodology based on the combination of the essential safety features and system parameter has been applied to the Zion nuclear power plant to evaluate the validity of the frontier utilized for the probabilistic safety assessment. The peak cladding temperature has been chosen as a relevant system parameter that represents the system behavior during the transient. The results showed that the frontier should be extended from 6 in. to 10 in. based on the required safety functions and system response.

  13. Application of the extended Kalman filtering for the estimation of core coolant flow rate in pressurized water reactors

    International Nuclear Information System (INIS)

    Shieh, D.J.; Upadhyaya, B.R.

    1986-01-01

    In-core neutron detector and core-exit temperature signals in a pressurized water reactor (PWR) satisfy the condition of observability of the core dynamic system, and can be used to estimate nonmeasurable state variables and model parameters. The extension of the Kalman filtering technique is very useful for direct parameter estimation. This approach is applied to the determination of core coolant mass flow rate in PWRs and is evaluated using in-core measurements at the Loss-of-Fluid Test (LOFT) reactor. The influence of model uncertainties on the estimation accuracy was studied using the ambiguity function analysis. A sequential discretization method was developed to achieve faster convergence to the true value, avoiding model discretization at each sample point. The performance of the extended Kalman filter and the computational innovations were evaluated using a reduced order core dynamic model of the LOFT reactor and random data simulation. The technique was then applied to the determination of LOFT core coolant flow rate from operational data at 100% and 65% flow conditions

  14. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  15. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  16. 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.

  17. The choice between two designs for the safety-injection system of a pressurized-water reactor, using probabilistic methods

    International Nuclear Information System (INIS)

    Villemeur, Alain

    1982-01-01

    A probabilistic study has been carried out to compare two designs for the safety-injection circuit of a pressurized-water reactor. It appears that unavailability of the circuit after an accident involving loss of coolant decreases little when one moves from a 2-line to a 3-line system. These results are compared with the disadvantages arising from increased redundancy, and in particular the increased cost of the installations. The 2-line circuit appears the optimum one on the basis of cost and reliability criteria. It has been chosen for the 1300-MWe units [fr

  18. Deuterium high pressure target

    International Nuclear Information System (INIS)

    Perevozchikov, V.V.; Yukhimchuk, A.A.; Vinogradov, Yu.I.

    2001-01-01

    The design of the deuterium high-pressure target is presented. The target having volume of 76 cm 3 serves to provide the experimental research of muon catalyzed fusion reactions in ultra-pure deuterium in the temperature range 80-800 K under pressures of up to 150 MPa. The operation of the main systems of the target is described: generation and purification of deuterium gas, refrigeration, heating, evacuation, automated control system and data collection system

  19. Assumptions used for evaluating the potential radiological consequences of a less of coolant accident for pressurized water reactors - June 1974

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility. The design basis loss of coolant accident is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety. This guide gives acceptable assumptions that may be used in evaluating the radiological consequences of this accident for a pressurized water reactor. In some cases, unusual site characteristics, plant design features, or other factors may require different assumptions which will be considered on an individual case basis. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position

  20. Residual-stresses in austenitic stainless-steel primary coolant pipes and welds of pressurized-water reactors

    International Nuclear Information System (INIS)

    Faure, F.; Leggatt, R.H.

    1996-01-01

    Surface and through thickness residual stress measurements were performed on an aged cast austenitic-ferritic stainless steel pipe and on an orbital TIG weld representative of those of primary coolant pipes in pressurized water reactors. An abrasive-jet hole drilling method and a block removal and layering method were used. Surface stresses and through thickness stress profiles are strongly dependent upon heat treatments, machining and welding operations. In the aged cast stainless steel pipe, stresses ranged between -250 and +175 MPa. On and near the orbital TIG weld, the outside surface of the weld was in tension both in the axial and hoop directions, with maximum values reaching 420 MPa in the weld. On the inside surface, the hoop stresses were compressive, reaching -300 MPa. However, the stresses in the axial direction at the root of the weld were tensile within 4 mm depth from the inside surface, locally reaching 280 MPa. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. Detection of stress corrosion cracks in reactor pressure vessel and primary coolant system anchor studs

    International Nuclear Information System (INIS)

    Light, G.M.; Joshi, N.R.

    1987-01-01

    Under Electric Power Research Institute (EPRI) contract No. 2179-2, southwest Research Institute is continuing work on the use of the cylindrically guided wave technique (CGWT) for inspecting stud bolts. Also being evaluated is the application of the CGWT to the inspection of reactor coolant pump shafts. Data have been collected for stud bolts ranging from 16 to 112 inches (40.6 to 285 cm) in length, and from 1 to 4.5 inches (2.54 to 11.4 cm) in diameter. For each bolt size, tests were conducted to determine the smallest detectable notch, the effect of thread noise, and the amount of detectable simulated corrosion. The ratio of reflected longitudinal signals to mode-converted signals was analyzed with respect to bolt diameter, bolt length, and frequency parameters. The results of these test showed the following: (1) The minimum detectable notch in the threaded region was approximately 0.05 inch (1.3 mm) for all stud bolts evaluated. (2) Thread noise could easily be detected, but the level of noise was below the minimum detectable notch signal. (3) For carbon steel, optimum transducer frequency was 5 MHz, using a transducer whose face had an impedance that matched the steel surface. (4) Simulated corrosion of 15% reduced diameter could be detected

  3. Feasibility of water injection into the turbine coolant to permit gas turbine contingency power for helicopter application

    Science.gov (United States)

    Vanfossen, G. J.

    1983-01-01

    A system which would allow a substantially increased output from a turboshaft engine for brief periods in emergency situations with little or no loss of turbine stress rupture life is proposed and studied analytically. The increased engine output is obtained by overtemperaturing the turbine; however, the temperature of the compressor bleed air used for hot section cooling is lowered by injecting and evaporating water. This decrease in cooling air temperature can offset the effect of increased gas temperature and increased shaft speed and thus keep turbine blade stress rupture life constant. The analysis utilized the NASA-Navy-Engine-Program or NNEP computer code to model the turboshaft engine in both design and off-design modes. This report is concerned with the effect of the proposed method of power augmentation on the engine cycle and turbine components. A simple cycle turboshaft engine with a 16:1 pressure ratio and a 1533 K (2760 R) turbine inlet temperature operating at sea level static conditions was studied to determine the possible power increase and the effect on turbine stress rupture life that could be expected using the proposed emergency cooling scheme. The analysis showed a 54 percent increse in output power can be achieved with no loss in gas generator turbine stress rupture life. A 231 K (415 F) rise in turbine inlet temperature is required for this level of augmentation. The required water flow rate was found to be .0109 kg water per kg of engine air flow.

  4. A summary of the assessment of fuel behaviour, fission product release and pressure tube integrity following a postulated large loss-of-coolant accident

    International Nuclear Information System (INIS)

    Langman, V.J.; Weaver, K.R.

    1984-05-01

    The Ontario Hydro analyses of fuel and pressure tube temperatures, fuel behaviour, fission product release and pressure tube integrity for large break loss-of-coolant accidents in Bruce A or Pickering A have been critically reviewed. The determinations of maximum fuel temperatures and fission product release are very uncertain, and pressure tube integrity cannot be assured where low steam flows are predicted to persist for times on the order of minutes

  5. The effect of zinc injection into PWR primary coolant on the reduction of radiation buildup and corrosion control. The solubilities of zinc, nickel and cobalt spinel oxides

    International Nuclear Information System (INIS)

    Miyajima, Kaori; Hirano, Hideo

    1999-01-01

    The use of zinc injection into PWR primary coolant to reduce radiation buildup has been widely studied, and te reduction effect has been experimentally confirmed. However, some items, such as the optimal concentration of zinc required to reduce radiation buildup, the corrosion control effect of zinc injection, and the influence of zinc injection on the integrity of fuel cladding, have not been clarified yet. In particular, the corrosion suppression effect of zinc remains unconfirmed. Therefore, it is necessary to measure and calculate the solubilities of zinc and nickel spinel oxides, which are formed on the surface of Ni-based alloys in PWR primary systems. In this study, in order to assess the effectiveness of zinc injection in the reduction of radiation buildup and the corrosion control of Ni-based alloy, the potential-pH diagrams for Zn-Cr-H 2 O, Ni-Cr-H 2 O, and Co-Cr-H 2 O systems at 300degC were constructed and the solubilities of Zn-Cr, Ni-Cr, and Co-Cr spinel oxides were calculated. It is concluded that under pH conditions for which NiCr 2 O 4 is stable, zinc injection is effective in corrosion control as well as in reducing radiation buildup. (author)

  6. The dynamic pressure measurements of the nuclear reactor coolant for condition-based maintenance of the reactor

    International Nuclear Information System (INIS)

    Es-Saheb, M.H.H.

    1990-01-01

    The condition-based maintenance of the nuclear reactor, by monitoring and measuring the instantaneous dynamic pressure distribution of the coolant (water) impact on the solid surfaces of the reactor during operation is presented. The behaviour of water domes (jets) produced by underwater explosions of small changes of P.E.T.N. at various depths in two different size cylindrical containers, which simulate the nuclear reactor, is investigated. Water surface domes (jets) from the underwater explosions are photographed. Depending on the depth of the charge, curved and flat top jets of up to 455 mm diameter and impact speeds of up to 70 m/sec. are observed. The instabilities in the dome surfaces are observed and the instantaneous profiles are analysed. It is found that, in all cases tested, the maximum pressure takes place at the center of the jet and could reach up to 3.0 times the on-dimensional impact pressure value. The use of their measurements, as online monitoring for condition-based maintenance and design-out maintenance is discussed. 18 refs

  7. The influence of slightly different main circulation pumps on PWR coolant pressure pulsations

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1989-01-01

    Pressure distribution along the core barrel circumference caused by the simultaneous operation of six main circulating pumps with slightly different revolutions obtained as a result of measurement in operated NPP is determined on the basis of the well-known Penzes method based on the solving of the wave equation with source term using the expansion into the infinite series of eigenfunctions. Results of calculations can be summarized as follows: the pressure distribution and the resulting force acting on the core barrel has a random character. The same is valid for core barrel vibrations and mainly for the joint between core barrel and pressure vessel. (orig.)

  8. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

  9. Application of the regulations on pressurized components or light water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Barthelemy, F.; Menjon, G.

    1977-01-01

    This paper describes the philosophy and the provisions of the Order of 26 February 1974 concerning application of the regulations on pressurized components for light water reactor steam supply systems. The aim is to show how these regulations which differ from other regulations on pressurized components and is more detailed on many points, is applied in practice in France in the various stages of the design, construction and operation of PWRs. (NEA) [fr

  10. High Pressure Biomass Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, Pradeep K [Georgia Tech Research Corporation, Atlanta, GA (United States)

    2016-07-29

    According to the Billion Ton Report, the U.S. has a large supply of biomass available that can supplement fossil fuels for producing chemicals and transportation fuels. Agricultural waste, forest residue, and energy crops offer potential benefits: renewable feedstock, zero to low CO2 emissions depending on the specific source, and domestic supply availability. Biomass can be converted into chemicals and fuels using one of several approaches: (i) biological platform converts corn into ethanol by using depolymerization of cellulose to form sugars followed by fermentation, (ii) low-temperature pyrolysis to obtain bio-oils which must be treated to reduce oxygen content via HDO hydrodeoxygenation), and (iii) high temperature pyrolysis to produce syngas (CO + H2). This last approach consists of producing syngas using the thermal platform which can be used to produce a variety of chemicals and fuels. The goal of this project was to develop an improved understanding of the gasification of biomass at high pressure conditions and how various gasification parameters might affect the gasification behavior. Since most downstream applications of synags conversion (e.g., alcohol synthesis, Fischer-Tropsch synthesis etc) involve utilizing high pressure catalytic processes, there is an interest in carrying out the biomass gasification at high pressure which can potentially reduce the gasifier size and subsequent downstream cleaning processes. It is traditionally accepted that high pressure should increase the gasification rates (kinetic effect). There is also precedence from coal gasification literature from the 1970s that high pressure gasification would be a beneficial route to consider. Traditional approach of using thermogravimetric analyzer (TGA) or high-pressure themogravimetric analyzer (PTGA) worked well in understanding the gasification kinetics of coal gasification which was useful in designing high pressure coal gasification processes. However

  11. Fascination at high pressures

    International Nuclear Information System (INIS)

    Chidambaram, R.

    1992-01-01

    Research at high pressures has developed into an interdisciplinary area which has important implications for and applications in the areas of physics, chemistry, materials sciences, planetary sciences, biology, engineering sciences and technology. The state of-the-art in this field is reviewed and future directions are indicated. (M.G.B.)

  12. Thrust Vectoring of a Continuous Rotating Detonation Engine by Changing the Local Injection Pressure

    International Nuclear Information System (INIS)

    Liu Shi-Jie; Lin Zhi-Yong; Sun Ming-Bo; Liu Wei-Dong

    2011-01-01

    The thrust vectoring ability of a continuous rotating detonation engine is numerically investigated, which is realized via increasing local injection stagnation pressure of half of the simulation domain compared to the other half. Under the homogeneous injection condition, both the flow-field structure and the detonation wave propagation process are analyzed. Due to the same injection condition along the inlet boundary, the outlines of fresh gas zones at different moments are similar to each other. The main flow-field features under thrust vectoring cases are similar to that under the baseline condition. However, due to the heterogeneous injection system, both the height of the fresh gas zone and the pressure value of the fresh gas in the high injection pressure zone are larger than that in the low injection pressure zone. Thus the average pressure in half of the engine is larger than that in the other half and the thrust vectoring adjustment is realized. (fundamental areas of phenomenology(including applications))

  13. Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

  14. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

  15. Pressure-temperature response of a full-pressure PWR containment to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Misak, J.

    1976-01-01

    A mathematical model and computer code TRACO III for pressure-temperature transients in the full-pressure containment of PWR during LOCA is described. Main attention is devoted to the analysis of parametric calculations with respect to the estimation of effect of various factors on the transient process and to the comparison of the theoretical and the experimental results on CVTR. (author)

  16. Injection Molding of High Aspect Ratio Nanostructures

    DEFF Research Database (Denmark)

    Matschuk, Maria; Larsen, Niels Bent

    We present a process for injection molding of 40 nm wide and >100 nm high pillars (pitch: 200 nm). We explored the effects of mold coatings and injection molding conditions on the replication quality of nanostructures in cyclic olefin copolymer. We found that optimization of molding parameters...

  17. Coolant cleanup method in a nuclear reactor

    International Nuclear Information System (INIS)

    Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

    1983-01-01

    Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

  18. Rapid isolation of biomarkers for compound specific radiocarbon dating using high-performance liquid chromatography and flow injection analysis-atmospheric pressure chemical ionisation mass spectrometry

    NARCIS (Netherlands)

    Sinninghe Damsté, J.S.; Smittenberg, R.H.; Hopmans, E.C.; Schouten, S.

    2002-01-01

    Repeated semi-preparative normal-phase HPLC was performed to isolate selected biomarkers from sediment extracts for radiocarbon analysis. Flow injection analysis mass spectrometry was used for rapid analysis of collected fractions to evaluate the separation procedure, taking only 1 min per fraction.

  19. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    International Nuclear Information System (INIS)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs

  20. Effects of temperature on corrosion fatigue crack growth of pressure vessel steels in PWR coolant

    International Nuclear Information System (INIS)

    Tice, D.R.; Bramwell, I.L.; Fairbrother, H.; Worswick, D.

    1994-01-01

    This paper presents experimental results concerning crack propagation rates in A508-III pressure vessel steel (medium sulphur content) exposed to PWR primary water at temperatures between 130 and 290 C. The results indicate that the greatest increase in corrosion fatigue crack growth rate occurs at temperatures in the range 150 to 200 C. Under these conditions, there was a marked change in the appearance of the fracture surface, with extensive micro-branching of the crack front and occasional bifurcation of the whole crack path. In contrast, at 290 C, the fracture surface is smoother, similar to that due to inert fatigue. The implication of these observations for assessment of the pressure vessel integrity, is examined. 14 refs., 15 figs., 3 tabs

  1. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  2. Effects of zinc injection on electrochemical corrosion and cracking behavior of stainless steels in borated and lithiated high temperature water

    International Nuclear Information System (INIS)

    Wu Xinqiang; Liu Xiahe; Han Enhou; Ke Wei

    2014-01-01

    Zinc (Zn) injection water chemistry (ZWC) adopted in primary coolant system in pressurized water reactors (PWRs) is to reduce the radiation buildup as well as retard the corrosion degradation in high temperature pressurized water through improving the characteristics of oxide scales formed on components materials. However, Zn injection involved corrosion and cracking behavior and related mechanisms are still under discussion. The understanding of Zn-bearing oxide scale characteristics and their protective property is of significance to clarify the environmentally assisted material failure problems in PWRs power plants. In the present work, in-situ potentiodynamic polarization curves and electrochemical impedance spectra measurements in high temperature borated and lithiated water as well as ex-situ X-ray photoelectron spectroscopy analyses have been done to investigate the effects of temperature (R.T.-603 K), pH T value at 573 K (6.9-7.4) and Zn-injection concentration (0-150 ppb) on electrochemical corrosion behavior and oxide scale characteristics of nuclear-grade stainless steels. The protective property of oxide scales under Zn-free and Zn-injected conditions degraded with increasing temperature, with Cr-rich oxide layer playing a key role on retarding further corrosion. The composition of oxide scales appeared slightly pH T dependent: rich in chromites and ferrites at pH T =6.9 and pH T =7.4, respectively. The corrosion rate decreased significantly in the high pH T value solution with Zn injection due to the formation of thin and compact oxide scales. The ≤50 ppb Zn injection could significantly affect the formation of Zn-bearing oxides on the surfaces, while >50 ppb Zn injection showed no obvious influence on the oxide scales. A modified point defect model was proposed to discuss the effects of injected Zn concentrations on the oxide scales in high temperature water. A 10 ppb Zn injection obviously decreased the intergranular cracking susceptibility of

  3. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  4. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    International Nuclear Information System (INIS)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

    2017-01-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  5. Cryogenic, Absolute, High Pressure Sensor

    Science.gov (United States)

    Chapman, John J. (Inventor); Shams. Qamar A. (Inventor); Powers, William T. (Inventor)

    2001-01-01

    A pressure sensor is provided for cryogenic, high pressure applications. A highly doped silicon piezoresistive pressure sensor is bonded to a silicon substrate in an absolute pressure sensing configuration. The absolute pressure sensor is bonded to an aluminum nitride substrate. Aluminum nitride has appropriate coefficient of thermal expansion for use with highly doped silicon at cryogenic temperatures. A group of sensors, either two sensors on two substrates or four sensors on a single substrate are packaged in a pressure vessel.

  6. Cryogenic High Pressure Sensor Module

    Science.gov (United States)

    Chapman, John J. (Inventor); Shams, Qamar A. (Inventor); Powers, William T. (Inventor)

    1999-01-01

    A pressure sensor is provided for cryogenic, high pressure applications. A highly doped silicon piezoresistive pressure sensor is bonded to a silicon substrate in an absolute pressure sensing configuration. The absolute pressure sensor is bonded to an aluminum nitride substrate. Aluminum nitride has appropriate coefficient of thermal expansion for use with highly doped silicon at cryogenic temperatures. A group of sensors, either two sensors on two substrates or four sensors on a single substrate are packaged in a pressure vessel.

  7. Contempt-LT: a computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Wheat, L.L.; Wagner, R.J.; Niederauer, G.F.; Obenchain, C.F.

    1975-06-01

    CONTEMPT-LT is a digital computer program, written in FORTRAN IV, developed to describe the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided to describe fan cooler and cooling spray engineered safety systems. Up to four compartments can be modeled with CONTEMPT-LT, and any compartment except the reactor system may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. CONTEMPT-LT can be used to model all current boiling water reactor pressure suppression systems, including containments with either vertical or horizontal vent systems. CONTEMPT-LT can also be used to model pressurized water reactor dry containments, subatmospheric containments, and dual volume containments with an annulus region, and can be used to describe containment responses in experimental containment systems. The program user defines which compartments are used, specifies input mass and energy additions, defines heat structure and leakage systems, and describes the time advancement and output control. CONTEMPT-LT source decks are available in double precision extended-binary-coded-decimal-interchange-code (EBCDIC) versions. Sample problems have been run on the IBM360/75 computer. (U.S.)

  8. SOCOOL-2, Molten Materials Na Coolant Interaction, Temperature and Pressure Transient

    International Nuclear Information System (INIS)

    Padilla, A. Jr.

    1973-01-01

    1 - Description of problem or function: SOCOOL2 calculates the transient temperatures, pressures, and mechanical work energy when a molten material is instantaneously and uniformly dispersed in liquid sodium which is initially under acoustic constraint. 2 - Method of solution: A unit cell consisting of a single spherical particle of molten material surrounded concentrically by sodium is used as the basis for the calculation. Heat transfer from the molten particle to the sodium is calculated by an implicit numerical technique assuming negligible contact resistance at the interface of the particle. The expansion of the heated sodium is calculated by the one-dimensional acoustic equation until vaporization conditions are attained. Upon vaporization, it is assumed that the particle becomes vapor-blanketed and that no further heat transfer to or from the sodium occurs. The heated sodium is then expanded to the specific final pressure in an isentropic expansion process. 3 - Restrictions on the complexity of the problem: The presence of an initial amount of sodium vapor or noncondensable gas cannot be taken into account. Time delays in the process of fragmentation and mixing of the molten material into the sodium cannot be considered. Heat transfer during the two-phase expansion of sodium is neglected

  9. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1975-01-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The analytical model used for the program is described. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc. 11 references. (U.S.)

  10. Severe accident in pressurized water reactors: molten fuel-coolant interaction

    International Nuclear Information System (INIS)

    Battail-Claret, Sylvie

    1993-01-01

    In order to study the phenomenon of interaction between corium and water, the author of this research thesis proposes a scenario to describe the behaviour of a drop of molten iron oxide suddenly plunged into a bath of liquid at room temperature. First, she addresses the modelling of the evolution of the vapour film which surrounds the hot drop and comprises a phase of establishment of a steady film and the phase of destabilisation of this film when an external pressure wave passes by. Besides, she modelled the process of fragmentation of a hot body induced by the destabilisation of a process due to the impact of liquid water micro-jets with water trapping in the hot body. Finally, a model of 'bubble dynamics' is proposed to describe the evolution of the vapour bubble fed by fragments. Theoretical results are compared with experimental results [fr

  11. Experimental study on cryogenic adsorption of methane by activated carbon for helium coolant purification of High-Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Chang, Hua; Wu, Zong-Xin; Jia, Hai-Jun

    2017-01-01

    Highlights: • The cryogenic CH 4 adsorption on activated carbon was studied for design of HTGR. • The breakthrough curves at different conditions were analyzed by the MTZ model. • The CH 4 adsorption isotherm was fitted well by the Toth model and the D-R model. • The work provides valuable reference data for helium coolant purification of HTGR. - Abstract: The cryogenic adsorption behavior of methane on activated carbon was investigated for helium coolant purification of high-temperature gas-cooled reactor by using dynamic column breakthrough method. With helium as carrier gas, experiments were performed at −196 °C and low methane partial pressure range of 0–120 Pa. The breakthrough curves at different superficial velocities and different feed concentrations were measured and analyzed by the mass-transfer zone model. The methane single-component adsorption isotherm was obtained and fitted well by the Toth model and the Dubinin-Radushkevich model. The adsorption heat of methane on activated carbon was estimated. The cryogenic adsorption process of methane on activated carbon has been verified to be effective for helium coolant purification of high-temperature gas-cooled reactor.

  12. 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)

  13. Recent results of three-dimensional CFD simulations of coolant mixing in VVER-440/213 reactor pressure vessel

    International Nuclear Information System (INIS)

    Kiss, B.; Boros, I.; Aszodi, A.

    2008-01-01

    The Budapest University of Technology and Economics, Institute of Nuclear Techniques has been working since 2001 on the three-dimensional CFD model of the reactor pressure vessel of the VVER-440 type reactor. During this time period - due to the development of the available computational capacity - a very complex and detailed model of the RPV has been developed. The aim of the construction of the new model is to describe further internal structures of the RPV (e.g. correct modeling of brake tubes, or internals in the upper mixing chamber) and to perform an extensive sensitivity analysis on the different modeling and calculation parameters (e.g. porous region models vs. detailed modeling, or n different turbulence models). The new model can be applied for steady state calculation during normal operational condition and for different transient analyses as well. One interesting application is the participation in a planned benchmark exercise on the start-up of the sixth main coolant pump, which is aimed to compare the capabilities of mixing models of one-dimensional system codes with the results of CFD simulation. (authors)

  14. BWR core response to fluctuations in coolant flow and pressure, with implications on noise diagnosis and stability monitoring

    International Nuclear Information System (INIS)

    Blomstrand, J.H.; Andersson, S.A.

    1982-01-01

    Reactor dynamic tests, utilizing sinuosidal oscillations in pressure and recirculation flow, have been conducted in operating BWRs in Sweden and Finland. Test data recorded, as well as recordings of process noise, have been analyzed in terms of dynamic core properties. The results obtained show good qualitative agreement with model predictions of BWR core dynamics. Model studies can often support interpretation of dynamic information obtained from operating plants. Comparisons between model studies, dynamic tests and process noise may also provide improved understanding of test results and noise patterns; in this way it can be demonstrated that some neutron flux noise is caused by noise in coolant flow and steam flow. From reactor test data nd noise recordings, core stability parameters have been evaluated by a number of methods. These have been found to provide essentially the same results. The cores investigated were found to be very stable under normal operating conditions. In special operating points, outside the normal operating range, higher decay ratios may occur. The experience indicates that for BWR cores, operated at decay ratios above quarter damping, the stability parameters may be identified from the oscillatory behavior of the autocorrelation in the time domain of the neutron flux noise

  15. Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

    2010-01-01

    For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

  16. The effect of nozzle diameter, injection pressure and ambient temperature on spray characteristics in diesel engine

    Science.gov (United States)

    Rhaodah Andsaler, Adiba; Khalid, Amir; Sharifhatul Adila Abdullah, Nor; Sapit, Azwan; Jaat, Norrizam

    2017-04-01

    Mixture formation of the ignition process is a key element in the diesel combustion as it influences the combustion process and exhaust emission. Aim of this study is to elucidate the effects of nozzle diameter, injection pressure and ambient temperature to the formation of spray. This study investigated diesel formation spray using Computational Fluid Dynamics. Multiphase volume of fluid (VOF) behaviour in the chamber are determined by means of transient simulation, Eulerian of two phases is used for implementation of mixing fuel and air. The detail behaviour of spray droplet diameter, spray penetration and spray breakup length was visualised using the ANSYS 16.1. This simulation was done in different nozzle diameter 0.12 mm and 0.2 mm performed at the ambient temperature 500 K and 700 K with different injection pressure 40 MPa, 70 MPa and 140 MPa. Results show that high pressure influence droplet diameter become smaller and the penetration length longer with the high injection pressure apply. Smaller nozzle diameter gives a shorter length of the breakup. It is necessary for nozzle diameter and ambient temperature condition to improve the formation of spray. High injection pressure is most effective in improvement of formation spray under higher ambient temperature and smaller nozzle diameter.

  17. Rapid isolation of biomarkers for compound specific radiocarbon dating using high-performance liquid chromatography and flow injection analysis-atmospheric pressure chemical ionisation mass spectrometry.

    Science.gov (United States)

    Smittenberg, Rienk H; Hopmans, Ellen C; Schouten, Stefan; Sinninghe Damsté, Jaap S

    2002-11-29

    Repeated semi-preparative normal-phase HPLC was performed to isolate selected biomarkers from sediment extracts for radiocarbon analysis. Flow injection analysis-mass spectrometry was used for rapid analysis of collected fractions to evaluate the separation procedure, taking only 1 min per fraction. In this way 100-1000 microg of glycerol dialkyl glycerol tetraethers, sterol fractions and chlorophyll-derived phytol were isolated from typically 100 g of marine sediment, i.e., in sufficient quantities for radiocarbon analysis, without significant carbon isotopic fractionation or contamination.

  18. Pressure Dome for High-Pressure Electrolyzer

    Science.gov (United States)

    Norman, Timothy; Schmitt, Edwin

    2012-01-01

    A high-strength, low-weight pressure vessel dome was designed specifically to house a high-pressure [2,000 psi (approx. = 13.8 MPa)] electrolyzer. In operation, the dome is filled with an inert gas pressurized to roughly 100 psi (approx. = 690 kPa) above the high, balanced pressure product oxygen and hydrogen gas streams. The inert gas acts to reduce the clamping load on electrolyzer stack tie bolts since the dome pressure acting axially inward helps offset the outward axial forces from the stack gas pressure. Likewise, radial and circumferential stresses on electrolyzer frames are minimized. Because the dome is operated at a higher pressure than the electrolyzer product gas, any external electrolyzer leak prevents oxygen or hydrogen from leaking into the dome. Instead the affected stack gas stream pressure rises detectably, thereby enabling a system shutdown. All electrical and fluid connections to the stack are made inside the pressure dome and require special plumbing and electrical dome interfaces for this to be accomplished. Further benefits of the dome are that it can act as a containment shield in the unlikely event of a catastrophic failure. Studies indicate that, for a given active area (and hence, cell ID), frame outside diameter must become ever larger to support stresses at higher operating pressures. This can lead to a large footprint and increased costs associated with thicker and/or larger diameter end-plates, tie-rods, and the frames themselves. One solution is to employ rings that fit snugly around the frame. This complicates stack assembly and is sometimes difficult to achieve in practice, as its success is strongly dependent on frame and ring tolerances, gas pressure, and operating temperature. A pressure dome permits an otherwise low-pressure stack to operate at higher pressures without growing the electrolyzer hardware. The pressure dome consists of two machined segments. An O-ring is placed in an O-ring groove in the flange of the bottom

  19. Vapour explosions (fuel-coolant interactions) resulting from the sub-surface injection of water into molten metals: preliminary results

    International Nuclear Information System (INIS)

    Asher, R.C.; Bullen, D.; Davies, D.

    1976-03-01

    Preliminary experiments are reported on the relationship between the injection mode of contact and the occurrence and magnitude of vapour explosions. Water was injected beneath the surface of molten metals, chiefly tin at 250 to 900 0 C. Vapour explosions occurred in many, but not all, cases. The results are compared with Dullforce's observations (Culham Report (CLM-P424) on the dropping mode of contact and it appears that rather different behaviour is found; in particular, the present results suggest that the Temperature Interaction Zone is different for the two modes of contact. (author)

  20. Neutronic performance of high molecular weight coolants for a prismatic VHTR

    International Nuclear Information System (INIS)

    Schriener, T. M.; El-Genk, M. S.

    2008-01-01

    A neutronic model is developed of a prismatic Very High Temperature Reactor (VHTR) to investigate the effects on the excess reactivity and operation cycle length of replacing helium with binary gas mixtures of He-Ne, He-N 2 , or He-Xe as reactor coolants and working fluids in the direct Closed Brayton Cycle (CBC) for energy conversion. Also investigated is the neutron activation of these binary gas mixtures in the VHTR. The motivation for using the heavy binary mixtures is the smaller size and the fewer number of stages of the CBC turbo-machinery. The present analysis uses the Monte Carlo code MCNPX 2.6D at typical operating conditions (500-1000 degrees and 7.12 MPa) in the VHTR. He-Ne (15 g/mol) is the best neutronically, but not thermal-hydraulically, followed by He-N 2 . Although He-Ne has ∼13.6% lower heat transfer coefficient than helium, it insignificantly affects the initial excess reactivity and the operation life cycle and experiences no neutrons activation. On the other hand, He-N 2 has 4.4% higher heat transfer coefficient than helium and experiences insignificant neutron activation in the reactor, but decreases the initial excess reactivity by ∼5.2% and the operation cycle length by 6.7%. He-Xe (15 g/mol) has 8% higher heat transfer coefficient than helium, but decreases the initial excess reactivity by 18.2% and the operational cycle length by 17%. In addition, neutron activation of xenon produces a significant source term, requiring shielding of the CBC loop and could contaminate the turbo-machinery with long-lived radioactive cesium. Thus, He-Xe is not recommended as a reactor coolant, but could be used as working fluid in a CBC loop that is indirectly coupled to helium cooled VHTR. (authors)

  1. In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

    International Nuclear Information System (INIS)

    Antonaccio, Carlos; Conde, Alberto; Fittipaldi, Andres H.; Maniotti, Jorge; Moliterno, Gabriel E.

    2000-01-01

    During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

  2. Numerical evaluation of various gas and coolant channel designs for high performance liquid-cooled proton exchange membrane fuel cell stacks

    International Nuclear Information System (INIS)

    Sasmito, Agus P.; Kurnia, Jundika C.; Mujumdar, Arun S.

    2012-01-01

    A careful design of gas and coolant channel is essential to ensure high performance and durability of proton exchange membrane (PEM) fuel cell stack. The channel design should allow for good thermal, water and gas management whilst keeping low pressure drop. This study evaluates numerically the performance of various gas and coolant channel designs simultaneously, e.g. parallel, serpentine, oblique-fins, coiled, parallel-serpentine and a novel hybrid parallel-serpentine-oblique-fins designs. The stack performance and local distributions of key parameters are investigated with regards to the thermal, water and gas management. The results indicate that the novel hybrid channel design yields the best performance as it constitutes to a lower pumping power and good thermal, water and gas management as compared to conventional channels. Advantages and limitation of the designs are discussed in the light of present numerical results. Finally, potential application and further improvement of the design are highlighted. -- Highlights: ► We evaluate various gas and coolant channel designs in liquid-cooled PEM fuel cell stack. ► The model considers coupled electrochemistry, channel design and cooling effect simultaneously. ► We propose a novel hybrid channel design. ► The novel hybrid channel design yields the best thermal, water and gas management which is beneficial for long term durability. ► The novel hybrid channel design exhibits the best performance.

  3. In reactor performance of defected zircaloy-clad U{sub 3}Si fuel elements in pressurized and boiling water coolants

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Allison, G M; Ambler, J F.R.; Chalder, G H; Lipsett, J J

    1968-05-15

    The results of two in-reactor defect tests of Zircaloy-clad U{sub 3}Si are reported. In the first test, a previously irradiated element ({approx}5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at {approx}270{sup o}C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U{sub 3}Si at 300{sup o}C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  4. Premixed direct injection nozzle for highly reactive fuels

    Science.gov (United States)

    Ziminsky, Willy Steve; Johnson, Thomas Edward; Lacy, Benjamin Paul; York, William David; Uhm, Jong Ho; Zuo, Baifang

    2013-09-24

    A fuel/air mixing tube for use in a fuel/air mixing tube bundle is provided. The fuel/air mixing tube includes an outer tube wall extending axially along a tube axis between an inlet end and an exit end, the outer tube wall having a thickness extending between an inner tube surface having a inner diameter and an outer tube surface having an outer tube diameter. The tube further includes at least one fuel injection hole having a fuel injection hole diameter extending through the outer tube wall, the fuel injection hole having an injection angle relative to the tube axis. The invention provides good fuel air mixing with low combustion generated NOx and low flow pressure loss translating to a high gas turbine efficiency, that is durable, and resistant to flame holding and flash back.

  5. The feasibility of water injection into the turbine coolant to permit gas turbine contingency power for helicopter application

    Science.gov (United States)

    Van Fossen, G. J.

    1983-01-01

    It is pointed out that in certain emergency situations it may be desirable to obtain power from a helicopter engine at levels greater than the maximum rating. Yost (1976) has reported studies concerning methods of power augmentation in the one engine inoperative (OEI) case. It was found that a combination of water/alcohol injection into the inlet and overtemperature/overspeed could provide adequate emergency power. The present investigation is concerned with the results of a feasibility study which analytically investigated the maximum possible level of augmentation with constant gas generator turbine stress rupture life as a constraint. In the proposed scheme, the increased engine output is obtained by turbine overtemperature, however, the temperature of the compressor bleed air used for hot section cooling is lowered by injecting and evaporating water.

  6. Material and water chemistry for a ferritic reactor coolant system in pressure water reactors

    International Nuclear Information System (INIS)

    Stieding, L.

    1979-04-01

    The use of unplated, low-alloy steels in a boric acid controlled PWR is not considered possible without changing the water conditions during the start-up and shut-down periods of the reactor. The significant pH reduction of the water due to boric acid during these periods most probably leads to damage of the magnetite protective layers followed by selective corrosion. As this highly important process has not been sufficiently evaluated with respect to our specific application problem, more detailed information will be necessary. KWU test facilities provide a means of performing such tests. In order to avoid corrosion attack during the above operating conditions, an inhibition of the water with 7 Li-borate is recommended which, however, will amount to approx. DM 60.000,-- per period of use. (orig.) [de

  7. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  8. High power neutral beam injection in LHD

    International Nuclear Information System (INIS)

    Tsumori, K.; Takeiri, Y.; Nagaoka, K.

    2005-01-01

    The results of high power injection with a neutral beam injection (NBI) system for the large helical device (LHD) are reported. The system consists of three beam-lines, and two hydrogen negative ion (H - ion) sources are installed in each beam-line. In order to improve the injection power, the new beam accelerator with multi-slot grounded grid (MSGG) has been developed and applied to one of the beam-lines. Using the accelerator, the maximum powers of 5.7 MW were achieved in 2003 and 2004, and the energy of 189 keV reached at maximum. The power and energy exceeded the design values of the individual beam-line for LHD. The other beam-lines also increased their injection power up to about 4 MW, and the total injection power of 13.1 MW was achieved with three beam-lines in 2003. Although the accelerator had an advantage in high power beam injection, it involved a demerit in the beam focal condition. The disadvantage was resolved by modifying the aperture shapes of the steering grid. (author)

  9. High pressure mechanical seal

    Science.gov (United States)

    Babel, Henry W. (Inventor); Anderson, Raymond H. (Inventor)

    1996-01-01

    A relatively impervious mechanical seal is formed between the outer surface of a tube and the inside surface of a mechanical fitting of a high pressure fluid or hydraulic system by applying a very thin soft metal layer onto the outer surface of the hard metal tube and/or inner surface of the hard metal fitting. The thickness of such thin metal layer is independent of the size of the tube and/or fittings. Many metals and alloys of those metals exhibit the requisite softness, including silver, gold, tin, platinum, indium, rhodium and cadmium. Suitably, the coating is about 0.0025 millimeters (0.10 mils) in thickness. After compression, the tube and fitting combination exhibits very low leak rates on the order or 10.sup.-8 cubic centimeters per second or less as measured using the Helium leak test.

  10. Common High Blood Pressure Myths

    Science.gov (United States)

    ... Disease Venous Thromboembolism Aortic Aneurysm More Common High Blood Pressure Myths Updated:May 4,2018 Knowing the facts ... This content was last reviewed October 2016. High Blood Pressure • Home • Get the Facts About HBP Introduction What ...

  11. Medications for High Blood Pressure

    Science.gov (United States)

    ... Consumers Home For Consumers Consumer Updates Medications for High Blood Pressure Share Tweet Linkedin Pin it More sharing options ... age and you cannot tell if you have high blood pressure by the way you feel, so have your ...

  12. High blood pressure and diet

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/007483.htm High blood pressure and diet To use the sharing features on ... diet is a proven way to help control high blood pressure . These changes can also help you lose weight ...

  13. Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

    International Nuclear Information System (INIS)

    Vigil, R.A.; Jacobus, M.J.

    1994-04-01

    Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables, the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent

  14. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  15. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    International Nuclear Information System (INIS)

    Moreira, Uebert G.; Dominguez, Dany S.

    2017-01-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  16. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    International Nuclear Information System (INIS)

    Sengpiel, W.

    1980-12-01

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW) [de

  17. Pressure and pressure derivative analysis for injection tests with variable temperature without type-curve matching

    International Nuclear Information System (INIS)

    Escobar, Freddy Humberto; Martinez, Javier Andres; Montealegre Matilde

    2008-01-01

    The analysis of injection tests under nonisothermic conditions is important for the accurate estimation of the reservoir permeability and the well's skin factor; since previously an isothermical system was assumed without taking into account a moving temperature front which expands with time plus the consequent changes in both viscosity and mobility between the cold and the hot zone of the reservoir which leads to unreliable estimation of the reservoir and well parameters. To construct the solution an analytical approach presented by Boughrara and Peres (2007) was used. That solution was initially introduced for the calculation of the injection pressure in an isothermic system. It was later modified by Boughrara and Reynolds (2007) to consider a system with variable temperature in vertical wells. In this work, the pressure response was obtained by numerical solution of the anisothermical model using the Gauss Quadrature method to solve the integrals, and assuming that both injection and reservoir temperatures were kept constant during the injection process and the water saturation is uniform throughout the reservoir. For interpretation purposes, a technique based upon the unique features of the pressure and pressure derivative curves were used without employing type-curve matching (TDS technique). The formulation was verified by its application to field and synthetic examples. As expected, increasing reservoir temperature causes a decrement in the mobility ratio, then estimation of reservoir permeability is some less accurate from the second radial flow, especially, as the mobility ratio increases

  18. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  19. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1976-06-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc

  20. Controlling your high blood pressure

    Science.gov (United States)

    ... medlineplus.gov/ency/patientinstructions/000101.htm Controlling your high blood pressure To use the sharing features on this page, ... JavaScript. Hypertension is another term used to describe high blood pressure. High blood pressure can lead to: Stroke Heart ...

  1. Superconductivity at high pressures

    Energy Technology Data Exchange (ETDEWEB)

    Brandt, N B; Ginzburg, N I

    1969-07-01

    Work published during the last 3 or 4 yrs concerning the effect of pressure on superconductivity is reviewed. Superconducting modifications of Si, Ge, Sb, Te, Se, P and Ce. Change of Fermi surface under pressure for nontransition metals. First experiments on the influence of pressure on the tunneling effect in superconductors provide new information on the nature of the change in phonon and electron energy spectra of metals under hydrostatic compression. 78 references.

  2. Computational Investigation of Novel Tip Leakage Mitigation Methods for High Pressure Turbine Blades

    Science.gov (United States)

    Ibrahim, Mounir; Gupta, Abhinav; Shyam, Vikram

    2014-01-01

    This paper presents preliminary findings on a possible approach to reducing tip leakage losses. In this paper a computational study was conducted on the Energy Efficient Engine (EEE) High Pressure Turbine (HPT) rotor tip geometry using the commercial numerical solver ANSYS FLUENT. The flow solver was validated against aerodynamic data acquired in the NASA Transonic Turbine Blade Cascade facility. The scope of the ongoing study is to computationally investigate how the tip leakage and overall blade losses are affected by (1) injection from the tip near the pressure side, (2) injection from the tip surface at the camber line, and (3) injection from the tip surface into the tip separation bubble. The objective is to identify the locations on the tip surface at which to place appropriately configured blowing keeping in mind the film cooling application of tip blowing holes. The validation was conducted at Reynolds numbers of 85,000, 343,000, and 685,000 and at engine realistic flow conditions. The coolant injection simulations were conducted at a Reynolds number of 343,000 based on blade chord and inlet velocity and utilized the SST turbulence model in FLUENT. The key parameters examined are the number of jets, jet angle and jet location. A coolant to inlet pressure ratio of 1.0 was studied for angles of +30 deg, -30 deg, and 90 deg to the local free stream on the tip. For the 3 hole configuration, 3 holes spaced 3 hole diameters apart with length to diameter ratio of 1.5 were used. A simulation including 11 holes along the entire mean camber line is also presented (30 deg toward suction side). In addition, the effect of a single hole is also compared to a flat tip with no injection. The results provide insight into tip flow control methods and can be used to guide further investigation into tip flow control. As noted in past research it is concluded that reducing leakage flow is not necessarily synonymous with reducing losses due to leakage.

  3. Coolant mixing in pressurized water reactors. Pt. 1. Feasibility of closed analytical solutions and simulation of the mixing with CFX-4. Final report

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Rohde, U.

    2001-10-01

    The project was aimed at the analytical and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing velocity measurements by the LDA technique. Design and construction of the ROCOM facility including the measurement equipment were performed in a second part of the project. For the design of the facility, CFD calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. A theoretical 2D-model of the downcomer flow was developed based on the potential theory. The coolant inlet is represented by mass sources. Potential vortices were superposed to describe large scale recirculations. However, the method requires an a-priory knowledge of the location and intensity of the vorticity sources. Therefore, the main goal of the project was the numerical simulation of the coolant mixing of different PWRs. The temperature and boron concentration fields established by the coolant mixing during nominal and transient flow conditions in the pressure vessel of the PWR Konvoi and the Russian type WWER-440 were investigated. The calculations were carried out with the CFD-code CFX 4. The results of the CFD calculation are found in the final report. The report is based on the Ph.D. work of T. Hoehne. (orig.) [de

  4. High-inertia hermetically sealed main coolant pump for next generation passive nuclear power plants

    International Nuclear Information System (INIS)

    Kujawski, Joseph M.; Nair, Bala R.; Vijuk, Ronald P.

    2003-01-01

    The main coolant pump for the Westinghouse AP1000 advanced passive nuclear power plant represents a significant scale-up in power, flow capacity, and physical size from its predecessor designed for the smaller AP600 power plant. More importantly, the AP1000 pump incorporates several innovative features that contribute to improved efficiency, operational reliability, and plant safety. The features include an internals design which provides the highest hydraulic efficiency achieved in commercial nuclear power plant applications. Another feature is the use of a distributed inertial mass system in the rotating assembly to develop the high rotational inertia to meet the extended system flow coastdown requirement for core heat removal in the event of loss of power to the pumps. This advanced canned motor pump also incorporates the latest development in higher operating voltage, providing plant designers with the ability to eliminate plant transformers and operate directly on the site electrical bus in many cases. The salient features of the pump design and performance data are presented in this paper. (author)

  5. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  6. High Blood Pressure in Pregnancy

    Science.gov (United States)

    ... of the baby. Controlling your blood pressure during pregnancy and getting regular prenatal care are important for ... your baby. Treatments for high blood pressure in pregnancy may include close monitoring of the baby, lifestyle ...

  7. Analysis of the loss of coolant accident due to the faiture in the open position of two pressurizer relief valves, for Angra-1 nuclear power plant

    International Nuclear Information System (INIS)

    Freire, C.F.

    1981-06-01

    A study of the modeling techniques adequate for simulating the loss of coolant accident caused by stuck open pressurizer relief valves, using the RELAP4-MOD5 code, is performed and the model developed is applied to the analysis of this kind of accident for the Central Nuclear Almirante Alvaro Alberto Unit (Angra 1). The thermal hydraulic behavior of the reactor cooling system, when subjected to a loss of main feedwater followed by the failure in the open position of two pressurizer relief valves, is determined. The relief valves are assumed to fail in the totally open position, delivering the maximum massflow through the discharge line. The RELAP4-MOD5 code is shown to be adequate for this kind of analysis, and the detailed prediction of the thermal hydraulic behavior of the Reactor Coolant System is thus possible. The eficiency of the emergency core cooling system of Angra 1 is demonstrated, the fuel elements remaining covered by the coolant during all the accident, and the peak clad temperatures are kept within design limites, ensuring the integrity of the core. (Author) [pt

  8. Breakdown pressures and characteristic flaw sizes during fluid injection experiments in shale at elevated confining pressures.

    Science.gov (United States)

    Chandler, M.; Mecklenburgh, J.; Rutter, E. H.; Taylor, R.; Fauchille, A. L.; Ma, L.; Lee, P. D.

    2017-12-01

    Fracture propagation trajectories in gas-bearing shales depend on the interaction between the anisotropic mechanical properties of the shale and the anisotropic in-situ stress field. However, there is a general paucity of available experimental data on their anisotropic mechanical, physical and fluid-flow properties, especially at elevated confining pressures. A suite of mechanical, flow and elastic measurements have been made on two shale materials, the Whitby mudrock and the Mancos shale (an interbedded silt and mudstone), as well as Pennant sandstone, an isotropic baseline and tight-gas sandstone analogue. Mechanical characterization includes standard triaxial experiments, pressure-dependent permeability, brazilian disk tensile strength, and fracture toughness determined using double-torsion experiments. Elastic characterisation was performed through ultrasonic velocities determined using a cross-correlation method. Additionally, we report the results of laboratory-scale fluid injection experiments for the same materials. Injection experiments involved the pressurisation of a blind-ending central hole in a dry cylindrical sample. Pressurisation is conducted under constant volume-rate control, using silicon oils of varying viscosities. Breakdown pressure is not seen to exhibit a strong dependence on rock type or orientation, and increases linearly with confining pressure. In most experiments, a small drop in the injection pressure record is observed at what is taken to be fracture initiation, and in the Pennant sandstone this is accompanied by a small burst of acoustic energy. The shale materials were acoustically quiet. Breakdown is found to be rapid and uncontrollable after initiation if injection is continued. A simplified 2-dimensional model for explaining this is presented in terms of the stress intensities at the tip of a pressurised crack, and is used alongside the triaxial data to derive a characteristic flaw size from which the fractures have initiated

  9. Gaseous poison injection device

    International Nuclear Information System (INIS)

    Kubota, Ryuji; Sugisaki, Toshihiko; Inada, Ikuo.

    1983-01-01

    Purpose: To rapidly control the chain reaction due to thermal neutrons in a reactor core by using gaseous poisons as back-up means for control rod drives. Constitution: Gaseous poisons having a large neutron absorption cross section are used as back-up means for control rod drives. Upon failure of control rod insertion, the gaseous poisons are injected into the lower portion of the reactor core to control the reactor power. As the gaseous poisons, vapors at a high temperature and a higher pressure than that of the coolants in the reactor core are injected to control the reactor power due to the void effects. Since the gaseous poisons thus employed rapidly reach the reactor core and form gas bubbles therein, the deccelerating effect of the thermal neutrons is decreased to reduce the chain reaction. (Moriyama, K.)

  10. Physical properties of organic coolants

    International Nuclear Information System (INIS)

    Debbage, A.G.; Garton, D.A.; Kinneir, J.H.

    1963-03-01

    Density, viscosity, specific heat, vapour pressure and calorific value were measured within the temperature range 100 - 400 deg C for mixtures of Santowax R with pyrolytic high boiler and Santowax R with O.M.R.E. radiolytic high boiler; in addition measurements were made on Santowax OM, X-7 standard, X-7 loop coolant and O.M.R.E. coolant supplied by Atomic Energy of Canada Ltd. The accuracy of the measurements made were density (± 1/4%), viscosity (± 2%), specific heat (± 2%), vapour pressure (± 2%) and calorific value (± 1/2%). Thermal conductivity was calculated from an improved form of the Smiths equation with an accuracy within ± 6%. Equations fitted to the vapour pressure results were used to provide data outside the experimental range for burnout correlation purposes. The general effect of high boiler content on the specific heat and calorific values was small. The differences in physical property values for corresponding values of either pyrolytic or radiolytic high boiler were small for density (0.3%) and specific heat (2%), but quite large for viscosity (70%) with the pyrolytic high boiler mixture giving the higher value. The chemical analysis of all materials was based on gas chromatography and the relationship between this and an earlier distillation method established. (author)

  11. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  12. Evaluation of responses to IE Bulletin 82-02: degradation of threaded fasteners in reactor coolant pressure boundary of pressurized-water-reactor plants

    International Nuclear Information System (INIS)

    Anderson, W.; Sterner, P.

    1985-05-01

    IE Bulletin 82-02 was issued by the NRC on June 2, 1982, to notify licensees about incidents of severe degradation of threaded fasteners. The bulletin required appropriate action including submittal of information from pressurized water reactors having an operating license. Responses from 41 licensees included their recent experience with degradation of threaded fasteners in primary system components. Data from recent regular inspections of reactor coolant pressure boundary component connections of 6-in. size and larger are compiled for technical evaluation. Statistical analysis is used to determine significant factors related to frequency of leakage incidents in connections, occurrence of degradation of bolts and studs, and the need for bolt replacement. Factors examined include the age of the plant, types of components, use of lubricants and sealants, and differences between plants. The compiled data indicate that, on the average, 10% of the bolted connections show evidence of leaking during an 18-month period. Also, 80% of the connections that show evidence of leakage undergo some degradation of the bolting. Results of the analysis show a significant decrease in the occurrence of bolting degradation events as the age of the plant increases. The data also show that valves are less subject to bolting corrosion. A group of 5 of the 41 plants accounted for about one-half of the reported leakage and corrosion events. The common characteristic found for four of these five plants was the lubricant used. The use of nickel-graphite based lubricants appears to offer a significantly reduced incidence of leakage and corrosion, based on late corrections to the reported data. The data also permit the conclusion that the use of molybdenum-disulfide-based lubricants and graphite-based lubricants results in a significantly increased incidence of leakage and corrosion. Reporting of data on lubricants was of poor quality and detracted from the value of the bulletin responses

  13. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  14. What Is High Blood Pressure?

    Science.gov (United States)

    ... Disease Venous Thromboembolism Aortic Aneurysm More What is High Blood Pressure? Updated:Feb 27,2018 First, let’s define high ... resources . This content was last reviewed October 2016. High Blood Pressure • Home • Get the Facts About HBP Introduction What ...

  15. High-temperature transient creep properties of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Chow, C.K.

    2002-06-01

    During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

  16. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1997-01-01

    This paper presents results of experimental studies on the heat transfer and solidifcation of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. As a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 .deg. C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleight number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer

  17. Small break LOCA [loss of coolant accident] mitigation for Bellefonte

    International Nuclear Information System (INIS)

    Bayless, P.D.; Dobbe, C.A.

    1986-01-01

    Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

  18. Contribution to the optimization of the chemical and radiochemical purification of pressurized water nuclear power plants primary coolant

    International Nuclear Information System (INIS)

    Elain, L.

    2004-12-01

    The primary coolant of pressurised water reactors is permanently purified thanks to a device, composed of filters and the demineralizers furnished with ion exchange resins (IER), located in the chemical and volume control system (CVCS). The study of the retention mechanisms of the radio-contaminants by the IER implies, initially, to know the speciation of the primary coolant percolant through the demineralizers. Calculations of theoretical speciation of the primary coolant were carried out on the basis of known composition of the primary coolant and thanks to the use of an adapted chemical speciation code. A complementary study, dedicated to silver behaviour, considered badly extracted, suggests metallic aggregates existence generated by the radiolytic reduction of the Ag + ions. An analysis of the purification curves of the elements Ni, Fe, Co, Cr, Mn, Sb and their principal radionuclides, relating to the cold shutdown of Fessenheim 1-cycle 20 and Tricastin 2-cycle 21, was carried out, in the light of a model based on the concept of a coupling well term - source term. Then, a thermodynamic modelling of ion exchange phenomena in column was established. The formation of the permutation front and the enrichment zones planned was validated by frontal analysis experiments of synthetic fluids (mixtures of Ni(B(OH) 4 ) 2 , LiB(OH) 4 and AgB(OH) 4 in medium B(OH) 3 )), and of real fluid during the putting into service of the device mini-CVCS at the time of Tricastin 2 cold shutdown. New tools are thus proposed, opening the way with an optimised management of demineralizers and a more complete interpretation of the available experience feedback. (author)

  19. Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Callow, R.A.

    1989-01-01

    The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

  20. High Blood Pressure - Multiple Languages

    Science.gov (United States)

    ... Being 8 - High Blood Pressure - Amarɨñña / አማርኛ (Amharic) MP3 Siloam Family Health Center Arabic (العربية) Expand Section ... Being 8 - High Blood Pressure - myanma bhasa (Burmese) MP3 Siloam Family Health Center Chinese, Simplified (Mandarin dialect) ( ...

  1. Analysis of a hot-leg small break loss-of-coolant accident in a three-loop westinghouse pressurized water reactor plant

    International Nuclear Information System (INIS)

    Peterson, C.E.; Chexal, V.K.; Clements, T.B.

    1985-01-01

    The RETRAN-02 computer code was used to perform a best-estimate analysis of a 7.52-cm-diam hotleg break in a three-loop Westinghouse pressurized water reactor. This break size produced a net primary coolant mass depletion through the early portion of the transient. The primary system started to refill only after the accumulator valves opened. As the primary system refilled, there were extreme temperature differentials around the system with cold, denser fluid collecting at the lower elevations and two-phase fluid at higher elevations

  2. Analysis of the behaviour of pressure and temperature of the containment of a PWR reactor, submitted to a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Silva, D.E. da; Arrieta, L.A.J.; Costa, J.R.; Camargo, C.; Santos, C.M. dos; Rochedo, E.R.R.

    1979-12-01

    The main purpose of this work is to analyse the pressure and temperature behaviour of the metalic containment of a PWR building, submitted to a postulated loss-of-coolant accident (LOCA) caused by a double-ended rupture in the main line of the primary circuit. The scope of the study was directed to verify the Final Safety Analysis Report (FSAR) results for the integrity of the metalic containment of the Angra I power plant. The highest containment pressure peak for this unit is expected for a break in the suction line of one of the main pumps of the primary coolant. Using the same input data, our results are very similar to those presented in the FSAR which shows a reasonable equivalence between the two analytical models. Using as input data the results of a previous LOCA study at CNEN, which yields to more conservative boundary conditions than those presented by the FSAR, the pressure and temperature peak values determined by our model are quite larger than those presented by the cited Safety Report. (author) [pt

  3. Fundamentals of high pressure adsorption

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Y.P.; Zhou, L. [Tianjin University, Tianjin (China). High Pressure Adsorption Laboratory

    2009-12-15

    High-pressure adsorption attracts research interests following the world's attention to alternative fuels, and it exerts essential effect on the study of hydrogen/methane storage and the development of novel materials addressing to the storage. However, theoretical puzzles in high-pressure adsorption hindered the progress of application studies. Therefore, the present paper addresses the major theoretical problems that challenged researchers: i.e., how to model the isotherms with maximum observed in high-pressure adsorption; what is the adsorption mechanism at high pressures; how do we determine the quantity of absolute adsorption based on experimental data. Ideology and methods to tackle these problems are elucidated, which lead to new insights into the nature of high-pressure adsorption and progress in application studies, for example, in modeling multicomponent adsorption, hydrogen storage, natural gas storage, and coalbed methane enrichment, was achieved.

  4. Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2017-01-01

    Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

  5. Definition of the seventh dynamic AER benchmark-WWER-440 pressure vessel coolant mixing by re-connection of an isolated loop

    International Nuclear Information System (INIS)

    Kotsarev, A.; Lizorkin, M.; Petrin, R.

    2010-01-01

    The seventh dynamic benchmark is a continuation of the efforts to validate systematically codes for the estimation of the transient behavior of VVER type nuclear power plants. This benchmark is a continuation of the work in the sixth dynamic benchmark. It is proposed to be simulated the transient - re-connection of an isolated circulating loop with low temperature or low boron concentration in a VVER-440 plant. It is supposed to expand the benchmark to other cases when a different number of loops are in operation leading to different symmetric and asymmetric core boundary conditions. The purposes of the proposed benchmark are: 1) Best-estimate simulations of an transient with a coolant flow mixing in the Reactor Pressure Vessel of WWER-440 plant by re-connection of one coolant loop to the several ones on operation, 2) Performing of code-to-code comparisons. The core is at the end of its first cycle with a power of 1196.25 MWt. The basic additional difference of the 7-seventh benchmark is in the detailed description of the downcomer and bottom part of the reactor vessel that allow describing the effects of coolant mixing in the Reactor Pressure Vessel without any additional conservative assumptions. The burn-up and the power distributions at this reactor state have to be calculated by the participants. The thermohydraulic conditions of the core in the beginning of the transient are specified. Participants self-generated best estimate nuclear data is to be used. The main geometrical parameters of the plant and the characteristics of the control and safety systems are also specified. Use generated input data decks developed for a WWER-440 plant and for the applied codes should be used. The behaviour of the plant should be studied applying coupled system codes, which combine a three-dimensional neutron kinetics description of the core with a pseudo or real 3D thermohydraulics system code. (Authors)

  6. Chemical aspects of hydrogen ingress in zirconium and zircaloy pressure tubes: ageing management of Indian PHWR coolant channels - determination of hydrogen and deuterium

    International Nuclear Information System (INIS)

    Sayi, Y.S.; Shankaran, P.S.; Yadav, C.S.; Ramanjaneyulu, P.S.; Venugopal, V.; Ramakumar, K.L.; Chhapru, G.C.; Prasad, R.; Jain, H.C.; Sood, D.D.

    2009-02-01

    Pressurized heavy water reactors (PHWRs) use zirconium and zirconium based alloys as clad and coolant tubes since its beginning. The first ever zircaloy-2 pressure tube failure occurred in 1983 at Ontario Hydro's Pickering Unit 2 in Canada which necessitated a thorough examination of causes of such failure. The failure was attributed to massive hydriding at the failed spot of pressure tube. Continuous usage of zirconium alloys could result in their hydrogen and deuterium pick-up leading to hydrogen/ deuterium embrittlement. The life of the zircaloy coolant channels is dictated by hydrogen/deuterium content and hence ageing management of the pressure tubes is essential for ensuring their trouble-free usage. It is desirable to have a sound knowledge on the chemical aspects of zirconium and zirconium based alloys metallurgy, the mechanistic principles of hydrogen ingress into the pressure tubes during in reactor service, and identifying suitable analytical methodologies for precise and accurate determination of hydrogen in wafer thin sliver samples carved out from insides of pressure tubes without causing any structural damage so that it can continue to remain in service. This is desirable so that the ageing management does not result in cost-escalation. This report is divided in to three main parts. The first part deals with the chemical aspects of zirconium and zirconium based alloy metallurgy, the mechanism of hydrogen pick-up and hydride formation in zirconium matrix. The second part describes various methodologies and their limitations, available for hydrogen/deuterium determination. The third part deals in detail, about the extensive investigations carried out at Radioanalytical Chemistry Division (RACD) in Radiochemistry and Isotope Group for establishing an indigenously developed hot vacuum extraction system in combination with quadrupole mass spectrometry for precise determination of hydrogen and deuterium in wafer thin sliver sample of zircaloy. The

  7. High pressure phase transformations revisited.

    Science.gov (United States)

    Levitas, Valery I

    2018-04-25

    High pressure phase transformations play an important role in the search for new materials and material synthesis, as well as in geophysics. However, they are poorly characterized, and phase transformation pressure and pressure hysteresis vary drastically in experiments of different researchers, with different pressure transmitting media, and with different material suppliers. Here we review the current state, challenges in studying phase transformations under high pressure, and the possible ways in overcoming the challenges. This field is critically compared with fields of phase transformations under normal pressure in steels and shape memory alloys, as well as plastic deformation of materials. The main reason for the above mentioned discrepancy is the lack of understanding that there is a fundamental difference between pressure-induced transformations under hydrostatic conditions, stress-induced transformations under nonhydrostatic conditions below yield, and strain-induced transformations during plastic flow. Each of these types of transformations has different mechanisms and requires a completely different thermodynamic and kinetic description and experimental characterization. In comparison with other fields the following challenges are indicated for high pressure phase transformation: (a) initial and evolving microstructure is not included in characterization of transformations; (b) continuum theory is poorly developed; (c) heterogeneous stress and strain fields in experiments are not determined, which leads to confusing material transformational properties with a system behavior. Some ways to advance the field of high pressure phase transformations are suggested. The key points are: (a) to take into account plastic deformations and microstructure evolution during transformations; (b) to formulate phase transformation criteria and kinetic equations in terms of stress and plastic strain tensors (instead of pressure alone); (c) to develop multiscale continuum

  8. High pressure phase transformations revisited

    Science.gov (United States)

    Levitas, Valery I.

    2018-04-01

    High pressure phase transformations play an important role in the search for new materials and material synthesis, as well as in geophysics. However, they are poorly characterized, and phase transformation pressure and pressure hysteresis vary drastically in experiments of different researchers, with different pressure transmitting media, and with different material suppliers. Here we review the current state, challenges in studying phase transformations under high pressure, and the possible ways in overcoming the challenges. This field is critically compared with fields of phase transformations under normal pressure in steels and shape memory alloys, as well as plastic deformation of materials. The main reason for the above mentioned discrepancy is the lack of understanding that there is a fundamental difference between pressure-induced transformations under hydrostatic conditions, stress-induced transformations under nonhydrostatic conditions below yield, and strain-induced transformations during plastic flow. Each of these types of transformations has different mechanisms and requires a completely different thermodynamic and kinetic description and experimental characterization. In comparison with other fields the following challenges are indicated for high pressure phase transformation: (a) initial and evolving microstructure is not included in characterization of transformations; (b) continuum theory is poorly developed; (c) heterogeneous stress and strain fields in experiments are not determined, which leads to confusing material transformational properties with a system behavior. Some ways to advance the field of high pressure phase transformations are suggested. The key points are: (a) to take into account plastic deformations and microstructure evolution during transformations; (b) to formulate phase transformation criteria and kinetic equations in terms of stress and plastic strain tensors (instead of pressure alone); (c) to develop multiscale continuum

  9. High pressure experimental water loop

    International Nuclear Information System (INIS)

    Grenon, M.

    1958-01-01

    A high pressure experimental water loop has been made for studying the detection and evolution of cladding failure in a pressurized reactor. The loop has been designed for a maximum temperature of 360 deg. C, a maximum of 160 kg/cm 2 and flow rates up to 5 m 3 /h. The entire loop consists of several parts: a main circuit with a canned rotor circulation pump, steam pressurizer, heating tubes, two hydro-cyclones (one de-gasser and one decanter) and one tubular heat exchanger; a continuous purification loop, connected in parallel, comprising pressure reducing valves and resin pots which also allow studies of the stability of resins under pressure, temperature and radiation; following the gas separator is a gas loop for studying the recombination of the radiolytic gases in the steam phase. The preceding circuits, as well as others, return to a low pressure storage circuit. The cold water of the low pressure storage flask is continuously reintroduced into the high pressure main circuit by means of a return pump at a maximum head of 160 kg /cm 2 , and adjusted to the pressurizer level. This loop is also a testing bench for the tight high pressure apparatus. The circulating pump and the connecting flanges (Oak Ridge type) are water-tight. The feed pump and the pressure reducing valves are not; the un-tight ones have a system of leak recovery. To permanently check the tightness the circuit has been fitted with a leak detection system (similar to the HRT one). (author) [fr

  10. Psoriasis and high blood pressure.

    Science.gov (United States)

    Salihbegovic, Eldina Malkic; Hadzigrahic, Nermina; Suljagic, Edin; Kurtalic, Nermina; Sadic, Sena; Zejcirovic, Alema; Mujacic, Almina

    2015-02-01

    Psoriasis is a chronic skin ailment which can be connected with an increased occurrence of other illnesses, including high blood pressure. A prospective study has been conducted which included 70 patients affected by psoriasis, both genders, older than 18 years. Average age being 47,14 (SD= ±15,41) years, from that there were 36 men or 51,43 and 34 women or 48,57%. Average duration of psoriasis was 15,52 (SD=±12,54) years. Frequency of high blood pressure in those affected by psoriasis was 54,28%. Average age of the patients with psoriasis and high blood pressure was 53,79 year (SD=±14,15) and average duration of psoriasis was 17,19 years (SD=±13,51). Average values of PASI score were 16,65. Increase in values of PASI score and high blood pressure were statistically highly related (r=0,36, p=0,0001). Psoriasis was related to high blood pressure and there was a correlation between the severity of psoriasis and high blood pressure.

  11. High Blood Pressure and Women

    Science.gov (United States)

    ... is known as gestational hypertension, a form of secondary hypertension caused by the pregnancy that usually disappears after delivery. If the mother is not treated, high blood pressure can be dangerous to both the mother ...

  12. High Pressure Industrial Water Facility

    Science.gov (United States)

    1992-01-01

    In conjunction with Space Shuttle Main Engine testing at Stennis, the Nordberg Water Pumps at the High Pressure Industrial Water Facility provide water for cooling the flame deflectors at the test stands during test firings.

  13. High Pressure Research on Materials

    Indian Academy of Sciences (India)

    example, represents the stress on the x plane in the y direction. There are three .... optical studies and studying compressibility of fluids. 3.2 Opposed ..... [4] G N Peggs, High Pressure Measurement Techniques, Applied Science. Publishers ...

  14. CONTEMPT-LT/028: a computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hargroves, D.W.; Metcalfe, L.J.; Wheat, L.L.; Niederauer, G.F.; Obenchain, C.F.

    1979-03-01

    CONTEMPT-LT is a digital computer program, written in FORTRAN IV, developed to describe the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided to describe fan cooler and cooling spray engineered safety systems. An annular fan model is also provided to model pressure control in the annular region of dual containment systems. Up to four compartments can be modeled with CONTEMPT-LT, and any compartment except the reactor system may have both a liquid pool region and an air--vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different

  15. Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

    2000-01-01

    The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

  16. High pressure gas reinjection unit

    Energy Technology Data Exchange (ETDEWEB)

    1976-03-01

    Nuovo Pignone has built for gas reinjection at Ekofisk the highest pressure injection unit to date: suction pressure 246 bar, discharge 647 bar, for 5.7 million cu m/day of natural gas, and driven by a GE MS 5001 gas turbine of 24,000 hp. The barrel-type compressor has been used already in Algeria at Hassi Messaoud. Full scale tests have shown that the unit is satisfactory; special attention being paid to the stability of the rotor. Air cooled heat exchangers were used in the test loop to cool the discharge gas; at Ekofisk, heat exchangers with sea water will be used. The valves in the test loop were of a special, low- noise type. Vibrations of the rotor system and changes in gas pressure monitored, showing that a pressure of 680 bars can be achieved without instability. Economic considerations lead to preference for rotary compressors driven by gas turbines for similar applications in the exploitation of oil fields. A graph of the characteristics of the unit is given.

  17. 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

  18. Reactor pressure vessel and reactor coolant circuit cast duplex stainless steel components contribution of the expertise for life management studies

    International Nuclear Information System (INIS)

    Bezdikian, Georges

    2006-09-01

    The life management of French Nuclear Power Plants is a major stake from an economic and a technical point of view considering the aging management assessment of the key components of the plant. The actual life evaluation is the result of prediction of life assessment from important program of expertise for the 3-loop PWR and 4-loop PWR plants in operation. To optimize the strategic policy in order to achieve the best possible performance and to prepare the technical and economical choice and decision, the paper presents the association of life management strategy and the program of expertise considering: - the identification of degradation for different components and prediction criteria proposed; - the large database from cast reactor coolant and component removed from nuclear power plants and expertise studies to confirm the prediction; - the life evaluation of RPV with radiation surveillance program based on the expertise of irradiation capsules, it is particularly shown how the expertise is in the center of the strategic choice. The French utility has organized the life management of nuclear plant as a function of several programs of expertise of knowledge on the long term experience feedback and the maintenance program for life. This paper shows updated on RPV and reactor coolant equipment activities engaged by utility on: - periodic maintenance and volume of expertise; - Alternative maintenance actions; - Large volume of expertise and how are managed these results to predict the aging management. (author)

  19. MDEP Technical Report TR-CSWG-03. Technical Report: fundamental attributes for the design and construction of reactor coolant pressure-boundary components

    International Nuclear Information System (INIS)

    2014-01-01

    The primary, long-term goal of MDEP's CSWG is to achieve international harmonisation of codes and standards for pressure boundary components in nuclear power plants that are important to reactor safety. The key to achieving harmonisation is to understand the extent of similarities and differences amongst the pressure boundary codes and standards used in various countries. To assist the CSWG in its long-term goals, several standards development organisations (SDOs) from various countries performed a comparison of their pressure boundary codes and standards to identify the extent of similarities and differences in code requirements and the reasons for their differences. This CSWG document provides the fundamental attributes which have been developed for the codes and standards used in the design and construction of reactor coolant pressure boundary components in nuclear power plants. The fundamental attributes are the basic concepts to be considered in the design, materials, fabrication, installation, examination, testing and over-pressure protection requirements for pressure boundary components

  20. Evaluation of containment peak pressure and structural response for a large-break loss-of-coolant accident in a VVER-440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W.; Sienicki, J.J.; Kulak, R.F.; Pfeiffer, P.A. [Argonne National Lab., IL (United States); Voeroess, L.; Techy, Z. [VEIKI Inst. for Electric Power Research, Budapest (Hungary); Katona, T. [Paks Nuclear Power Plant (Hungary)

    1998-07-01

    A collaborative effort between US and Hungarian specialists was undertaken to investigate the response of a VVER-440/213-type NPP to a maximum design-basis accident, defined as a guillotine rupture with double-ended flow from the largest pipe (500 mm) in the reactor coolant system. Analyses were performed to evaluate the magnitude of the peak containment pressure and temperature for this event; additional analyses were performed to evaluate the ultimate strength capability of the containment. Separate cases were evaluated assuming 100% effectiveness of the bubbler-condenser pressure suppression system as well as zero effectiveness. The pipe break energy release conditions were evaluated from three sources: (1) FSAR release rate based on Soviet safety calculations, (2) RETRAN-03 analysis and (3) ATHLET analysis. The findings indicated that for 100% bubbler-condenser effectiveness the peak containment pressures were less than the containment design pressure of 0.25 MPa. For the BDBA case of zero effectiveness of the bubbler-condenser system, the peak pressures were less than the calculated containment failure pressure of 0.40 MPa absolute.

  1. Pressure thermal shock analysis for nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

    2015-01-01

    The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

  2. 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)

  3. High-Compression-Ratio; Atkinson-Cycle Engine Using Low-Pressure Direct Injection and Pneumatic-Electronic Valve Actuation Enabled by Ionization Current and Foward-Backward Mass Air Flow Sensor Feedback

    Energy Technology Data Exchange (ETDEWEB)

    Harold Schock; Farhad Jaberi; Ahmed Naguib; Guoming Zhu; David Hung

    2007-12-31

    This report describes the work completed over a two and one half year effort sponsored by the US Department of Energy. The goal was to demonstrate the technology needed to produce a highly efficient engine enabled by several technologies which were to be developed in the course of the work. The technologies included: (1) A low-pressure direct injection system; (2) A mass air flow sensor which would measure the net airflow into the engine on a per cycle basis; (3) A feedback control system enabled by measuring ionization current signals from the spark plug gap; and (4) An infinitely variable cam actuation system based on a pneumatic-hydraulic valve actuation These developments were supplemented by the use of advanced large eddy simulations as well as evaluations of fuel air mixing using the KIVA and WAVE models. The simulations were accompanied by experimental verification when possible. In this effort a solid base has been established for continued development of the advanced engine concepts originally proposed. Due to problems with the valve actuation system a complete demonstration of the engine concept originally proposed was not possible. Some of the highlights that were accomplished during this effort are: (1) A forward-backward mass air flow sensor has been developed and a patent application for the device has been submitted. We are optimistic that this technology will have a particular application in variable valve timing direct injection systems for IC engines. (2) The biggest effort on this project has involved the development of the pneumatic-hydraulic valve actuation system. This system was originally purchased from Cargine, a Swedish supplier and is in the development stage. To date we have not been able to use the actuators to control the exhaust valves, although the actuators have been successfully employed to control the intake valves. The reason for this is the additional complication associated with variable back pressure on the exhaust valves when

  4. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  5. Experimental investigation of the effects of blowing conditions and Mach number on the unsteady behavior of coolant ejection through a trailing edge cutback

    International Nuclear Information System (INIS)

    Barigozzi, Giovanna; Armellini, Alessandro; Mucignat, Claudio; Casarsa, Luca

    2012-01-01

    Highlights: ► Flow visualization and PIV documented the presence of large coherent structures. ► The presence of coherent structures is documented up to the vane trailing edge. ► Shape and direction of rotation of vortices change with injection conditions. ► Vortices morphology influences the film cooling effectiveness distributions. ► A Mach number increase moves vortices closer to the wall. - Abstract: The present paper shows the results of an experimental investigation into the unsteadiness of coolant ejection at the trailing edge of a highly loaded nozzle vane cascade. The trailing edge cooling scheme features a pressure side cutback with film cooling slots, stiffened by evenly spaced ribs in an inline configuration. Cooling air is also ejected through two rows of cylindrical holes placed upstream of the cutback. Tests were performed with a low inlet turbulence intensity level (Tu 1 = 1.6%), changing the cascade operating conditions from low speed (M 2is = 0.2) up to high subsonic regime (M 2is = 0.6), and with coolant to main stream mass flow ratio varied within the 0.5–2.0% range. Particle Image Velocimetry (PIV) and flow visualizations were used to investigate the unsteady mixing process taking place between coolant and main flow downstream of the cutback, up to the trailing edge. For all the tested conditions, the results show the presence of large coherent structures, which presence is still evident up to the trailing edge. Their shape and direction of rotation change with injection conditions, as a function of coolant to mainstream velocity ratio, strongly influencing the thermal protection capability of the injected coolant flow. The Mach number increase is only responsible for a positioning of such vortical structures closer to the wall, while the Strouhal number almost remains unchanged.

  6. Zinc injection on the EDF pressurized light water reactors. Current results and operating experience feedback

    International Nuclear Information System (INIS)

    Piana, Olivier; Duval, Arnaud; Moleiro, Edgar; Benfarah, Moez; Bretelle, Jean-Luc; Chaigne, Guy

    2014-01-01

    Nowadays, zinc injection, as well as pH management and hydrogen control, is increasingly considered as an essential element of PWR Primary Water Chemistry worldwide. After a first implementation of zinc injection at Bugey 2 since 2004 and Bugey 4 since 2006, EDF decided to extend this practice, which constitutes a modification of primary circuit chemical conditioning, to other units of its fleet. Currently, 15 among the 58 reactors of the French fleet are injecting depleted zinc acetate into the primary coolant water. Three main goals were identified at the beginning of this program. Indeed, the expected benefits of zinc injection were: Reduction of the rate of generalized corrosion and mitigation of stress corrosion cracking initiation on nickel based alloys (Material goal). Curative or preventive reduction of radiation sources to which workers are exposed (Radiation fields' goal). Mitigation of the AOA or CIPS risks by reduction of corrosion products releases and mitigation of crud deposition (Fuel protection goal). To monitor the zinc addition, EDF has defined a complete survey program concerning: chemistry and radiochemistry responses (primary coolant monitoring of corrosion and fission products and calculation of zinc injected, zinc removed and zinc incorporated in RCS surfaces) ; radiation fields (dose rates and deposited activities measurements) ; materials (statistical analysis of SG tube cracks) ; fuel (oxide thickness measurements and visual exams) ; effluents (corrosion products releases and isotopic distribution follow up) ; wastes (radiochemical characterization of filters). This paper will detail the present results of this monitoring program. It appears that the expected benefits of zinc injection have yet to be fully realized; further operating experience will be required in order to fully evaluate its impact. (author)

  7. Thermodynamic data for selected gas impurities in the primary coolant of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Feber, R.C.

    1976-12-01

    The literature of thermodynamic data for selected fission-product species is reviewed and supplemented in support of complex chemical equilibrium calculations applied to fission-product distributions in the primary coolant of high-temperature gas-cooled reactors. Thermodynamic functions and heats and free energies of formation are calculated and tabulated to 3000 0 K for CsI (s,l,g), Cs 2 I 2 (g), CH 3 I(g), COI 2 (g), and CsH(g). 79 references

  8. Cavitation and primary atomization in real injectors at low injection pressure condition

    Science.gov (United States)

    Dumouchel, Christophe; Leboucher, Nicolas; Lisiecki, Denis

    2013-06-01

    This experimental work investigates the influence of the geometry of GDI devices on primary atomization processes under low injection pressure and reduced back pressure. These pressure conditions ensure cavitating flows and observable atomization processes. Measurements include mass flux, structure velocity from high-speed visualizations and spray characterization with a laser diffraction technique. Super-cavitation regime and cavitation string, which have their own influence on the mass flux, develop independently in different injector regions. These regimes impact the flow pattern in the orifice and the subsequent atomization process. A possible interaction between cavitation string and super-cavitation is found to promote a hydraulic-flip-like regime and to deteriorate atomization quality. As far as the geometry of the injector is concerned, the profile of the orifice inlet and the roughness of the sac volume region are found to be important geometrical characteristics.

  9. Compressed air injection technique to standardize block injection pressures : [La technique d'injection d'air comprimé pour normaliser les pressions d'injection d'un blocage nerveux].

    Science.gov (United States)

    Tsui, Ban C H; Li, Lisa X Y; Pillay, Jennifer J

    2006-11-01

    Presently, no standardized technique exists to monitor injection pressures during peripheral nerve blocks. Our objective was to determine if a compressed air injection technique, using an in vitro model based on Boyle's law and typical regional anesthesia equipment, could consistently maintain injection pressures below a 1293 mmHg level associated with clinically significant nerve injury. Injection pressures for 20 and 30 mL syringes with various needle sizes ( 18G, 20G, 21 G, 22G, and 24G) were measured in a closed system. A set volume of air was aspirated into a saline-filled syringe and then compressed and maintained at various percentages while pressure was measured. The needle was inserted into the injection port of a pressure sensor, which had attached extension tubing with an injection plug clamped "off". Using linear regression with all data points, the pressure value and 99% confidence interval (CI) at 50% air compression was estimated. The linearity of Boyle's law was demonstrated with a high correlation, r = 0.99, and a slope of 0.984 (99% CI: 0.967-1.001). The net pressure generated at 50% compression was estimated as 744.8 mmHg, with the 99% CI between 729.6 and 760.0 mmHg. The various syringe/needle combinations had similar results. By creating and maintaining syringe air compression at 50% or less, injection pressures will be substantially below the 1293 mmHg threshold considered to be an associated risk factor for clinically significant nerve injury. This technique may allow simple, real-time and objective monitoring during local anesthetic injections while inherently reducing injection speed. Présentement, aucune technique normalisée ne permet de vérifier les pressions d'injection pendant les blocages nerveux périphériques. Nous voulions vérifier si une technique d'injection d'air comprimé, utilisant un modèle in vitro fondé sur la loi de Boyle et du matériel propre à l'anesthésie régionale, pouvait maintenir avec régularité les

  10. High Blood Pressure (Hypertension) (For Parents)

    Science.gov (United States)

    ... Safe Videos for Educators Search English Español Hypertension (High Blood Pressure) KidsHealth / For Parents / Hypertension (High Blood Pressure) What's ... High Blood Pressure) Treated? Print What Is Hypertension (High Blood Pressure)? Blood pressure is the pressure of blood against ...

  11. Determination of boron as boric acid by automatic potentiometric titration using Gran plots [in pressurized water reactor coolant

    International Nuclear Information System (INIS)

    Midgley, D.; Gatford, C.

    1989-11-01

    Boron in PWR primary coolant and related waters may be determined as boric acid by titration with sodium hydroxide, using a glass electrode as a pH indicator. Earlier work has shown that this analysis can conveniently be carried out automatically with adequate precision and accuracy for routine use, although bias became apparent at the lowest concentrations tested. The latest titrators enable the titration data to be transformed mathematically to give two linear segments, before and after the end-point (Gran plots). The results are as precise as those from other titration methods (in which the end-point is found from the point of inflexion of a plot of pH against volume of titrant), but the bias at low concentrations is much reduced. This is achieved without extra time or involvement of the operator. (author)

  12. Effect of Intra Vitreal Injection of Bevacizumab on Intra-Occular Pressure

    International Nuclear Information System (INIS)

    Jaffar, S.; Tayyab, A.; Matin, Z. I.; Masrur, A.; Naqaish, R.

    2016-01-01

    Background: Bevacizumab has been in use as a therapeutic agent for macular oedema for several years. While its efficacy has been well documented, its use has been shown to cause a transient rise in the intra-ocular pressure. The aim of this study was to evaluate the long term effect of intra-vitreal injection of Bevacizumab on Intra-ocular pressure. Methods: One hundred eyes (n=100) of one hundred patients, requiring intra-vitreal injection of Bevacizumab for diabetic macular oedema were recruited from Shifa Foundation Community Health Centre (SFCHC) between January and December 2014. Patients of glaucoma, ocular hyper-tension, known allergy to Bevacizumab or had injections of Bevacizumab prior to the study were excluded. Intra-ocular pressure was measured using a Goldmann applanation tonometer, prior to, and at six and twelve months after the injection. The pre- and post- injection Intra-ocular pressure was entered into the database. Test of significance was applied to investigate whether there was a significant change in intra-ocular pressure after the injection. Results: The mean age of the patient was 56.97 years (±14.97). The mean intra-ocular pressure was 13.86 (±3.16) mmHg before injection, while post-injection mean Intra-Ocular pressure was 14.21 (±3.12) mmHg and 13.79 (±3.07) at six and twelve months respectively. Between baseline and six months there was a statistically significant difference in intra-ocular pressure (p=0.03), while no significant difference existed in the intra-ocular pressure between baseline and twelve months (p=0.92). Conclusion: Intra-vitreal injection of Bevacizumab is associated with a statically significant rise in intra-ocular pressure at six months, while no significant difference was seen at twelve months compared to baseline. (author)

  13. Terbium oxide at high pressures

    International Nuclear Information System (INIS)

    Dogra, Sugandha; Sharma, Nita Dilawar; Singh, Jasveer; Bandhyopadhyay, A.K.

    2011-01-01

    In this work we report the behaviour of terbium oxide at high pressures. The as received sample was characterized at ambient by X-ray diffraction and Raman spectroscopy. The X-ray diffraction showed the sample to be predominantly cubic Tb 4 O 7 , although a few peaks also match closely with Tb 2 O 3 . In fact in a recent study done on the same sample, the sample has been shown to be a mixture of Tb 4 O 7 and Tb 2 O 3 . The sample was subjected to high pressures using a Mao-Bell type diamond anvil cell upto a pressure of about 42 GPa with ruby as pressure monitor

  14. Decoupling Analysis on Pressure Fluctuation and Needle Valve Response for High Pressure Common Rail Injector

    Directory of Open Access Journals (Sweden)

    Hao Wang

    2017-01-01

    Full Text Available In the process of multiple injections, the influence of different injections makes the controlling of cycle fuel injection quantity more difficult. The high pressure common rail (HPCR simulation model is established in AMESim environment. Through the method of combining numerical simulation and experiment test, it is found that the strong coupling of pressure fluctuation and needle valve response is the fundamental reason, which leads to the fluctuation of main injection fuel quantity (MIFQ with dwell time (DT. The result shows that the largest fluctuation quantity is 3.6mm3 when the reference value of main injection is 60.0mm3. Non-damping LC hydraulic system model is also established. Through the analysis of the model, reducing the length-diameter ratio of internal oil duct and the delivery chamber volume are decoupling methods to the strong coupling.

  15. High Pressure Treatment in Foods

    OpenAIRE

    Edwin Fabian Torres Bello; Gerardo González Martínez; Bernadette F. Klotz Ceberio; Dolores Rodrigo; Antonio Martínez López

    2014-01-01

    Abstract: High hydrostatic pressure (HHP), a non-thermal technology, which typically uses water as a pressure transfer medium, is characterized by a minimal impact on food characteristics (sensory, nutritional, and functional). Today, this technology, present in many food companies, can effectively inactivate bacterial cells and many enzymes. All this makes HHP very attractive, with very good acceptance by consumers, who value the organoleptic characteristics of products processed by this non...

  16. Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Accidental loss of pool coolant due to leakage in a PWR SFP was studied using MAAP5. • The effect of emergency ventilation on the accident progression was investigated. • The effect of emergency injection on the accident progression was discussed. - Abstract: A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent

  17. Solubilities of iron and nickel oxides under high temperature and high pressure conditions

    International Nuclear Information System (INIS)

    Choi, Ke-Chon; Jung, Yong-Ju; Yeon, Jei-Won; Jee, Kwang-Yong

    2007-01-01

    The purposes of primary coolant chemistry are to assure fuel and material integrity and to minimize out of core radiation fields. During the PWR operation, crud deposits are expected on the cladding, leading to cladding failure and raising the radioactivity. Such deposits come from the corrosion products of system surface. To achieve optimal conditions for primary coolant, basic researches on mass transfer, deposition and solubility of corrosion products are needed. The initial stage of crud formation could be the studies on the solubility of a structural material. It has been known that the solubility of metal oxides in boric acid under high temperature and high pressure condition depends on the pH and dissolved hydrogen. Thus, the effect of various pH on the solubility of metal oxide in boric acid solution was investigated in this work

  18. Transient effects caused by pulsed gas and liquid injections into low pressure plasmas

    International Nuclear Information System (INIS)

    Ogawa, D; Goeckner, M; Overzet, L; Chung, C W

    2010-01-01

    The fast injection of liquid droplets into a glow discharge causes significant time variations in the pressure, the chemical composition of the gas and the phases present (liquid and/or solid along with gas). While the variations can be large and important, very few studies, especially kinetic studies, have been published. In this paper we examine the changes brought about in argon plasma by injecting Ar (gas), N 2 (gas) hexane (gas) and hexane (liquid droplets). The changes in the RF capacitively coupled power (forward and reflected), electron and ion density (n e , n i ), electron temperature (T e ) and optical emissions were monitored during the injections. It was found that the Ar injection (pressure change only) caused expected variations. The electron temperature reduced, the plasma density increased and the optical emission intensity remained nearly constant. The N 2 and hexane gas injections (chemical composition and pressure changes) also followed expected trends. The plasma densities increased and electron temperature decreased while the optical emissions changed from argon to the injected gas. These all serve to highlight the fact that the injection of evaporating hexane droplets in the plasma caused very little change. This is because the number of injected droplets is too small to noticeably affect the plasma, even though the shift in the chemical composition of the gas caused by evaporation from those same droplets can be very significant. The net conclusion is that using liquid droplets to inject precursors for low pressure plasmas is both feasible and controllable.

  19. Coolant Passage

    Directory of Open Access Journals (Sweden)

    Tom I.-P. Shih

    2001-01-01

    Full Text Available Computations were performed to study the three-dimensional flow and heat transfer in a U-shaped duct of square cross section with inclined ribs on two opposite walls under rotating and non-rotating conditions. Two extreme limits in the Reynolds number (25,000 and 350,000 were investigated. The rotation numbers investigated are 0, 0.24, and 0.039. Results show rotation and the bend to reinforce secondary flows that align with it and to retard those that do not. Rotation was found to affect significantly the flow and heat transfer in the bend even at a very high Reynolds number of 350,000 and a very low Rotation number of 0:039. When there is no rotation, the flow and heat transfer in the bend were dominated by rib-induced secondary flows at the high Reynolds number limit and by bend-induced pressure-gradients at the low Reynolds number limit. Long streaks of reduced surface heat transfer occur in the bend at locations where streamlines from two contiguous secondary flows merge and then flow away from the surface. The location and size of these streaks varied markedly with Reynolds and rotation numbers.

  20. Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Alekseev, P. N.; Semchenkov, Yu. M.; Sedov, A. A.; Subbotin, S. A.; Chibinyaev, A. V.

    2011-01-01

    Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.

  1. An on-line pressurizer surveillance system design to prevent small-break loss-of-coolant accidents through power-operated relief valves using a microcomputer

    International Nuclear Information System (INIS)

    Lee, J.H.; Chang, S.H.

    1987-01-01

    A small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve is one of the important contributors to nuclear power plant risk. A pressurizer surveillance system was designed to use a microcomputer to prevent the malfunction of the system; the effect of this improvement has been assessed through probabilistic risk assessment. The microcomputer diagnoses the malfunction of the system by a process-checking method and automatically performs the backup action related to each malfunction. This improvement means that we can correctly diagnose ''spurious opening,'' ''failure to reclose,'' and ''small-break LOCA,'' which are difficult for operators to diagnose quickly and correctly, and by taking automatic backup action one can reduce the probability of human error

  2. 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)

  3. Pressure vessel failure at high internal pressure

    International Nuclear Information System (INIS)

    Laemmer, H.; Ritter, B.

    1995-01-01

    A RPV failure due to plastic instability was investigated using the ABAQUS finite element code together with a material model of thermal plasticity for large deformations. Not only rotational symmetric temperature distributions were studied, but also 'hot spots'. Calculations show that merely by the depletion of strength of the material - even at internal wall temperatures well below the melting point of the fuel elements of about 2000/2400 C - the critical internal pressure can decrease to values smaller than the operational pressure of 16 Mpa. (orig.)

  4. Radiation pressure injection in laser-wakefield acceleration

    Science.gov (United States)

    Liu, Y. L.; Kuramitsu, Y.; Isayama, S.; Chen, S. H.

    2018-01-01

    We investigated the injection of electrons in laser-wakefield acceleration induced by a self-modulated laser pulse by a two dimensional particle-in-cell simulation. The localized electric fields and magnetic fields are excited by the counter-streaming flows on the surface of the ion bubble, owing to the Weibel or two stream like instability. The electrons are injected into the ion bubble from the sides of it and then accelerated by the wakefield. Contrary to the conventional wave breaking model, the injection of monoenergetic electrons are mainly caused by the electromagnetic process. A simple model was proposed to address the instability, and the growth rate was verified numerically and theoretically.

  5. The development of robotic system for inspecting and repairing NPP primary coolant system of high-level radioactive environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Kim, Ki Ho; Jung, Seung Ho; Kim, Byung Soo; Hwang, Suk Yeoung; Kim, Chang Hoi; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Cho, Jai Wan; Lee, Jae Kyung; Lee, Yong Deok

    1997-07-01

    This project aims at developing a robotic system to automatically handle inspection and maintenance of NPP safety-related facilities in high-level radioactive environment. This robotic system under development comprises two robots depending on application fields - a mobile robot and multi-functional robot. The mobile robot is designed to be used in the area of primary coolant system during the operation of NPP. This robot enables to overcome obstacles and perform specified tasks in unstructured environment. The multi-functional robot is designed for performing inspection and maintenance tasks of steam generator and nuclear reactor vessel during the overhaul periods of NPP. Nuclear facilities can be inspected and repaired all the time by use of both the mobile robot and the multi-functional robot. Human operator, by teleoperation, monitors the movements of such robots located at remote task environment via video cameras and controls those remotely generating desired commands via master manipulator. We summarize the technology relating to the application of the mobile robot to primary coolant system environment, the applicability of the mobile robot through 3D graphic simulation, the design of the mobile robot, the design of its radiation-hardened controller. We also describe the mechanical design, modeling, and control system of the multi-functional robot. Finally, we present the design of the force-reflecting master and the modeling of virtual task environment for a training simulator. (author). 47 refs., 16 tabs., 43 figs.

  6. The development of robotic system for inspecting and repairing NPP primary coolant system of high-level radioactive environment

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Kim, Ki Ho; Jung, Seung Ho; Kim, Byung Soo; Hwang, Suk Yeoung; Kim, Chang Hoi; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Cho, Jai Wan; Lee, Jae Kyung; Lee, Yong Deok.

    1997-07-01

    This project aims at developing a robotic system to automatically handle inspection and maintenance of NPP safety-related facilities in high-level radioactive environment. This robotic system under development comprises two robots depending on application fields - a mobile robot and multi-functional robot. The mobile robot is designed to be used in the area of primary coolant system during the operation of NPP. This robot enables to overcome obstacles and perform specified tasks in unstructured environment. The multi-functional robot is designed for performing inspection and maintenance tasks of steam generator and nuclear reactor vessel during the overhaul periods of NPP. Nuclear facilities can be inspected and repaired all the time by use of both the mobile robot and the multi-functional robot. Human operator, by teleoperation, monitors the movements of such robots located at remote task environment via video cameras and controls those remotely generating desired commands via master manipulator. We summarize the technology relating to the application of the mobile robot to primary coolant system environment, the applicability of the mobile robot through 3D graphic simulation, the design of the mobile robot, the design of its radiation-hardened controller. We also describe the mechanical design, modeling, and control system of the multi-functional robot. Finally, we present the design of the force-reflecting master and the modeling of virtual task environment for a training simulator. (author). 47 refs., 16 tabs., 43 figs

  7. Heat and momentum transfer in a gas coolant flow through a circular pipe in a high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Ogawa, Masuro

    1989-07-01

    In Japan Atomic Energy Research Institute (JAERI), a very high temperature gas cooled reactor (VHTR) has been researched and developed with a purpose of attaining a coolant temperature of around 1000degC at the reactor outlet. In order to design VHTR, comprehensive knowledge is required on thermo-hydraulic characteristics of laminar-turbulent transition, of coolant flow with large thermal property variation due to temperature difference, and of heat transfer deterioration. In the present investigation, experimental and analytical studies are made on a gas flow in a circular tube to elucidate the thermo-hydraulic characteristics. Friction factors and heat transfer coefficients in transitional flows are obtained. Influence of thermal property variation on the friction factor is qualitatively determined. Heat transfer deterioration in the turbulent flow subjected to intense heating is experimentally found to be caused by flow laminarization. The analysis based on a k-kL two-equation model of turbulence predicts well the experimental results on friction factors and heat transfer coefficients in flows with thermal property variation and in laminarizing flows. (author)

  8. Coupled analysis of passive safety injection and containment filtered venting for passive decay heat removal - 15140

    International Nuclear Information System (INIS)

    Kim, S.H.; Ham, J.H.; Jeong, Y.H.; Chang, S.H.

    2015-01-01

    Lots of interests for the safety of nuclear power plants have risen these days. The safety has to be continuously reviewed and enhanced in nuclear power plants currently operating as well as those designed and constructed in future. After the Fukushima accidents, many additional safety systems which can be applied to nuclear power plants in operation have been proposed. Those include alternating power source such as movable diesel generators and DC batteries in non-safety grade. Also, emergency preparedness for the prevention of a core damage accident was proposed to cope with the extended-SBO (station blackout) by using fire protection systems. In order to prevent the release of radioactive materials, safety systems for preserving the integrity of containment were proposed in two views of cooling and venting containment. Two approaches are effective for mitigating a severe accident. The design concept installing big water tanks besides containment at high level was proposed for various safety functions. One of the functions in the system is to inject the coolant from the elevated tank into a reactor vessel in the case of loss of coolant accident. When the pressure in reactor coolant system is sufficiently low, the coolant can be injected by gravity. If not, the depressurization in reactor vessel would be needed considering the containment pressure. Containment cooling in conventional pressurized water reactors is dependent on containment cooling pumps and sprays. Additional containment cooling systems cannot be simply and easily applied in the current nuclear power plants without major modifications. Therefore, for the operation of passive safety injection system, containment filtered venting system can be adopted for the depressurization of containment. In the design and operation of the passive safety injection system and the containment filtered venting system, main operating points related with open and close pressures in the filtered venting system were

  9. Computer simulation at high pressure

    International Nuclear Information System (INIS)

    Alder, B.J.

    1977-11-01

    The use of either the Monte Carlo or molecular dynamics method to generate equations-of-state data for various materials at high pressure is discussed. Particular emphasis is given to phase diagrams, such as the generation of various types of critical lines for mixtures, melting, structural and electronic transitions in solids, two-phase ionic fluid systems of astrophysical interest, as well as a brief aside of possible eutectic behavior in the interior of the earth. Then the application of the molecular dynamics method to predict transport coefficients and the neutron scattering function is discussed with a view as to what special features high pressure brings out. Lastly, an analysis by these computational methods of the measured intensity and frequency spectrum of depolarized light and also of the deviation of the dielectric measurements from the constancy of the Clausius--Mosotti function is given that leads to predictions of how the electronic structure of an atom distorts with pressure

  10. Anxiety: A Cause of High Blood Pressure?

    Science.gov (United States)

    ... of high blood pressure? Can anxiety cause high blood pressure? Answers from Sheldon G. Sheps, M.D. Anxiety doesn't cause long-term high blood pressure (hypertension). But episodes of anxiety can cause dramatic, ...

  11. African Americans and High Blood Pressure

    Science.gov (United States)

    ANSWERS by heart Lifestyle + Risk Reduction High Blood Pressure What About African Americans and High Blood Pressure? African Americans in the U.S. have a higher prevalence of high blood pressure (HBP) than ...

  12. High Blood Pressure: Medicines to Help You

    Science.gov (United States)

    ... For Consumers Consumer Information by Audience For Women High Blood Pressure--Medicines to Help You Share Tweet Linkedin Pin ... Click here for the Color Version (PDF 533KB) High blood pressure is a serious illness. High blood pressure is ...

  13. The behaviour of zirconium alloys in Santowax OM organic coolant at high temperatures

    International Nuclear Information System (INIS)

    Sawatzky, A.

    1964-10-01

    Zirconium alloys have been exposed to Santowax OM at temperatures of 320 to 400 o C for times as long as 5000 hours. Short-term experiments (less than 2 weeks) were done in stainless-steel bombs and small out-of-pile loops. The X-7 organic loop in the NRX reactor was used to study long-term oxidation and hydriding both in-flux and out-of-flux. The results obtained lead to several tentative conclusions: Aluminum cladding serves as an effective hydrogen barrier; Considerable protection against hydriding is given by zirconium oxide, provided impurities in the organic are carefully controlled; Hydriding is greatly enhanced by the presence of chlorine in the coolant; and, Hydriding is somewhat enhanced by neutron irradiation. Of considerable significance is the fact that a Zircaloy-4 in-reactor test section of the X-7 loop was exposed to Santowax OM at 320 to 400 o C for more than 5000 hours without excessive hydriding. (author)

  14. High-pressure sodium lamp

    NARCIS (Netherlands)

    1996-01-01

    A high pressure sodium lamp of the invention is provided with a discharge vessel (20) which is enclosed with intervening space (1) by an outer bulb (10), which space contains a gas-fill with at least 70 mol. % nitrogen gas. Electrodes (30a, 30b) are positioned in the discharge vessel (20) and are

  15. Intermolecular Interactions at high pressure

    DEFF Research Database (Denmark)

    Eikeland, Espen Zink

    2016-01-01

    In this project high-pressure single crystal X-ray diffraction has been combined with quantitative energy calculations to probe the energy landscape of three hydroquinone clathrates enclosing different guest molecules. The simplicity of the hydroquinone clathrate structures together with their st...

  16. High-pressure water facility

    Science.gov (United States)

    2006-01-01

    NASA Test Operations Group employees, from left, Todd Pearson, Tim Delcuze and Rodney Wilkinson maintain a water pump in Stennis Space Center's high-pressure water facility. The three were part of a group of employees who rode out Hurricane Katrina at the facility and helped protect NASA's rocket engine test complex.

  17. Vital Signs - High Blood Pressure

    Centers for Disease Control (CDC) Podcasts

    2012-10-02

    In the U.S., nearly one third of the adult population have high blood pressure, the leading risk factor for heart disease and stroke - two of the nation's leading causes of death.  Created: 10/2/2012 by National Center for Chronic Disease Prevention and Health Promotion (NCCDPHP).   Date Released: 10/17/2012.

  18. A reavaluation of the reliability analysis of the low pressure injection system for Angra-1

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Fleming, P.V.; Frutuoso e Melo, P.F.F.; Tayt-Sohn, L.C.

    1983-01-01

    The emergency core cooling system of Angra 1 is analysed aiming at the low pressure injection systems, using the fault tree technique. All the failure mode of the components are considered for this analyse. (author) [pt

  19. High Pressure Treatment in Foods

    Science.gov (United States)

    Torres Bello, Edwin Fabian; González Martínez, Gerardo; Klotz Ceberio, Bernadette F.; Rodrigo, Dolores; Martínez López, Antonio

    2014-01-01

    High hydrostatic pressure (HHP), a non-thermal technology, which typically uses water as a pressure transfer medium, is characterized by a minimal impact on food characteristics (sensory, nutritional, and functional). Today, this technology, present in many food companies, can effectively inactivate bacterial cells and many enzymes. All this makes HHP very attractive, with very good acceptance by consumers, who value the organoleptic characteristics of products processed by this non-thermal food preservation technology because they associate these products with fresh-like. On the other hand, this technology reduces the need for non-natural synthetic additives of low consumer acceptance. PMID:28234332

  20. High Pressure Treatment in Foods

    Directory of Open Access Journals (Sweden)

    Edwin Fabian Torres Bello

    2014-08-01

    Full Text Available High hydrostatic pressure (HHP, a non-thermal technology, which typically uses water as a pressure transfer medium, is characterized by a minimal impact on food characteristics (sensory, nutritional, and functional. Today, this technology, present in many food companies, can effectively inactivate bacterial cells and many enzymes. All this makes HHP very attractive, with very good acceptance by consumers, who value the organoleptic characteristics of products processed by this non-thermal food preservation technology because they associate these products with fresh-like. On the other hand, this technology reduces the need for non-natural synthetic additives of low consumer acceptance.

  1. Phase transitions in solids under high pressure

    CERN Document Server

    Blank, Vladimir Davydovich

    2013-01-01

    Phase equilibria and kinetics of phase transformations under high pressureEquipment and methods for the study of phase transformations in solids at high pressuresPhase transformations of carbon and boron nitride at high pressure and deformation under pressurePhase transitions in Si and Ge at high pressure and deformation under pressurePolymorphic α-ω transformation in titanium, zirconium and zirconium-titanium alloys Phase transformations in iron and its alloys at high pressure Phase transformations in gallium and ceriumOn the possible polymorphic transformations in transition metals under pressurePressure-induced polymorphic transformations in АIBVII compoundsPhase transformations in AIIBVI and AIIIBV semiconductor compoundsEffect of pressure on the kinetics of phase transformations in iron alloysTransformations during deformation at high pressure Effects due to phase transformations at high pressureKinetics and hysteresis in high-temperature polymorphic transformations under pressureHysteresis and kineti...

  2. Possibility of a pressurized water reactor concept with highly inherent heat removal following capability

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Murao, Yoshio

    1995-01-01

    If the core power inherently follows change in heat removal rate from the primary coolant system within small thermal expansion of the coolant which can be absorbed in a practical size of pressurizer, reactor systems may have more safety and load following capability. In order to know possibility and necessary conditions of a concept on reactor core and primary coolant system of a pressurized water reactor (PWR) with such 'highly inherent heat removal following capability', transient analyses on an ordinary two-loop PWR have been performed for a transient due to 50% change in heat removal with the RETRAN-02 code. The possibility of a PWR concept with the highly inherent heat removal following capability has been demonstrated under the conditions of the absolute value of ratio of the coolant density reactivity coefficient to the Doppler reactivity coefficient more than 10x10 3 kg·cm 3 which is two to three times larger than that at beginning of cycle (BOC) in an ordinary PWR and realized by elimination of the chemical shim, the 12% lower average linear heat generation rate of 17.9 kW/m, and the 1.5 times larger pressurizer volume than those of the ordinary PWR. (author)

  3. Metallic sodium as a coolant of high speed nuclear reactors, (2)

    International Nuclear Information System (INIS)

    Atsumo, Hideo

    1975-01-01

    Tables are given on all the sodium loops in Japan and most of the sodium loops all over the world. Name and purpose of the loops, time of establishment, highest temperature, amount of sodium, flow rate, the materials used for the construction of the loops, and the diameter of the main pipings are given. Also, the problems related with these loops are discussed. For example, the high temperature sodium facility at HEDL-WADCO was made for the FFTF component test and instrument test, and uses 50,000 gallons of metallic sodium. The highest temperature is 590 0 C. The sodium flows at the rate of 60 g/m. The body is made of Type 304 stainless steel. Main data of existing sodium-cooled reactors in the world are also tabulated. The data include thermal output, electric output, the structure of the reactor cores, the dimensions of the cores, fuel used, the highest temperature in the reactors, the temperature of sodium at the inlet and outlet, the rate of multiplication, the amount of sodium used, number of control rods, number of heat exchangers, and the pressure of steam. The Monju type nuclear reactor in Japan uses 1,800 ton of sodium. (Fukutomi, T.)

  4. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    International Nuclear Information System (INIS)

    Kalkahoran, Omid Noori; Ahangari, Rohollah; Shirani, Amir Saied

    2016-01-01

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results

  5. Simulation of Containment Pressurization in a Large Break-Loss of Coolant Accident Using Single-Cell and Multicell Models and CONTAIN Code

    Directory of Open Access Journals (Sweden)

    Omid Noori-Kalkhoran

    2016-10-01

    Full Text Available Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model. In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code’s results.

  6. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Kalkahoran, Omid Noori; Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

  7. Experiment data report for LOFT large-break loss-of-coolant experiment L2-5

    International Nuclear Information System (INIS)

    Bayless, P.D.; Divine, J.M.

    1982-08-01

    Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)] commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected

  8. Primary coolant feed and bleed operating regions for the Midland Plant

    International Nuclear Information System (INIS)

    Tsai, M.S.

    1985-01-01

    Operating regions for primary coolant feed and bleed cooling are developed for the Midland Plant using core decay heat, the high-pressure injection (HPI) system capacity, and flow rate relief through the power-operated relief valve (PORV). This mode of cooling is used for accident scenarios in which the normal core cooling means of a nuclear power plant is lost because of loss of water inventory in the steam generators. The HPI flow is based on the capacities of one and two pumps. Saturated steam, saturated water, and subcooled water are considered to be possible states of the fluid being relieved through the PORV. In estimating the PORV relief rate, flow equations are derived from the Electric Power Research Institute test data obtained from the same model and size valve that is used in the Midland Plant. For easy reference by operators, the operating region is displayed on a plane of reactor coolant system pressure and temperature. The technique developed for the Midland Plant provides a convenient method for examining the feed and bleed cooling capability for a nuclear power plant that employs a pressurized water reactor system

  9. Large Eddy Simulation of Cryogenic Injection Processes at Supercritical Pressure

    Science.gov (United States)

    Oefelein, Joseph C.

    2002-01-01

    This paper highlights results from the first of a series of hierarchical simulations aimed at assessing the modeling requirements for application of the large eddy simulation technique to cryogenic injection and combustion processes in liquid rocket engines. The focus is on liquid-oxygen-hydrogen coaxial injectors at a condition where the liquid-oxygen is injected at a subcritical temperature into a supercritical environment. For this situation a diffusion dominated mode of combustion occurs in the presence of exceedingly large thermophysical property gradients. Though continuous, these gradients approach the behavior of a contact discontinuity. Significant real gas effects and transport anomalies coexist locally in colder regions of the flow, with ideal gas and transport characteristics occurring within the flame zone. The current focal point is on the interfacial region between the liquid-oxygen core and the coaxial hydrogen jet where the flame anchors itself.

  10. High Blood Pressure: Unique to Older Adults

    Science.gov (United States)

    ... our e-newsletter! Aging & Health A to Z High Blood Pressure Hypertension Unique to Older Adults This section provides ... Pressure Targets are Different for Very Old Adults High blood pressure (also called hypertension) increases your chance of having ...

  11. Transient heating effects in high pressure Diesel injector nozzles

    International Nuclear Information System (INIS)

    Strotos, George; Koukouvinis, Phoevos; Theodorakakos, Andreas; Gavaises, Manolis; Bergeles, George

    2015-01-01

    Highlights: • Simulation of friction-induced heating in high pressure Diesel fuel injectors. • Injection pressures up to 3000 bar. • Simulations with variable fuel properties significantly affect predictions. • Needle motion affects flow and temperature fields. • Possible heterogeneous boiling as injection pressures increase above 2000 bar. - Abstract: The tendency of today’s fuel injection systems to reach injection pressures up to 3000 bar in order to meet forthcoming emission regulations may significantly increase liquid temperatures due to friction heating; this paper identifies numerically the importance of fuel pressurization, phase-change due to cavitation, wall heat transfer and needle valve motion on the fluid heating induced in high pressure Diesel fuel injectors. These parameters affect the nozzle discharge coefficient (C d ), fuel exit temperature, cavitation volume fraction and temperature distribution within the nozzle. Variable fuel properties, being a function of the local pressure and temperature are found necessary in order to simulate accurately the effects of depressurization and heating induced by friction forces. Comparison of CFD predictions against a 0-D thermodynamic model, indicates that although the mean exit temperature increase relative to the initial fuel temperature is proportional to (1 − C d 2 ) at fixed needle positions, it can significantly deviate from this value when the motion of the needle valve, controlling the opening and closing of the injection process, is taken into consideration. Increasing the inlet pressure from 2000 bar, which is the pressure utilized in today’s fuel systems to 3000 bar, results to significantly increased fluid temperatures above the boiling point of the Diesel fuel components and therefore regions of potential heterogeneous fuel boiling are identified

  12. Experimental Study of Injection Characteristics of a Multi-hole port injector on various Fuel Injection pressures and Temperatures

    Directory of Open Access Journals (Sweden)

    Ommi F

    2013-04-01

    Full Text Available The structures of the port injector spray dominates the mixture preparation process and strongly affect the subsequent engine combustion characteristics over a wide range of operating conditions in port-injection gasoline engines. All these spray characteristics are determined by particular injector design and operating conditions. In this paper, an experimental study is made to characterize the breakup mechanism and spray characteristics of a injector with multi-disc nozzle (SAGEM,D2159MA. A comparison was made on injection characteristics of the multi-hole injectors and its effects on various fuel pressure and temperature. The distributions of the droplet size and velocity and volume flux were characterized using phase Doppler anemometry (PDA technique. Through this work, it was found that the injector produces a finer spray with a wide spray angle in higher fuel pressure and temperature.

  13. Experimental Study of Injection Characteristics of a Multi-hole port injector on various Fuel Injection pressures and Temperatures

    Science.gov (United States)

    Movahednejad, E.; Ommi, F.; Nekofar, K.

    2013-04-01

    The structures of the port injector spray dominates the mixture preparation process and strongly affect the subsequent engine combustion characteristics over a wide range of operating conditions in port-injection gasoline engines. All these spray characteristics are determined by particular injector design and operating conditions. In this paper, an experimental study is made to characterize the breakup mechanism and spray characteristics of a injector with multi-disc nozzle (SAGEM,D2159MA). A comparison was made on injection characteristics of the multi-hole injectors and its effects on various fuel pressure and temperature. The distributions of the droplet size and velocity and volume flux were characterized using phase Doppler anemometry (PDA) technique. Through this work, it was found that the injector produces a finer spray with a wide spray angle in higher fuel pressure and temperature.

  14. The behaviour of CAGR moderator and sleeve graphites radiolytically oxidised to high weight loss in inhibited coolant gas compositions

    International Nuclear Information System (INIS)

    Schofield, P.; Fitzgerald, B.; Ketchen, J.

    1987-01-01

    Gilsocarbon graphites were irradiated to high weight losses in three different CO 2 based coolants. The experimental data is tested against a model which interprets the gas phase chemistry and pore geometry and allows weight loss and gas flow properties to be calculated. The observed changes of oxidation rate with dose were successfully predicted from the model. An empirical relationship was also derived which was shown to fit data for moderator, sleeve and special pore structure graphites. Changes in graphite permeability and diffusivity were predicted by the model, and also by other simplified, more approximate methods. The model based upon the measured transport pore spectrum was shown to be the best with other methods proving adequate to moderate doses. (author)

  15. Primary coolant chemistry of the Peach Bottom and Fort St. Vrain high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Burnette, R.D.; Baldwin, N.L.

    1980-11-01

    The chemical impurities in the primary coolants of the Peach Bottom and Fort St. Vrain reactors are discussed. The impurity mixtures in the two plants were quite different because the sources of the impurities were different. In the Peach Bottom reactor, the impurities were dominated by H 2 and CH 4 , which are decomposition products of oil. In the Fort St. Vrain reactor, there were high levels of CO, CO 2 , and H 2 O. Although oil ingress at Peach Bottom created carbon deposits on virtually all surfaces, its effect on reactor operation was negligible. Slow outgassing of water from the thermal insulation at Fort St. Vrain caused delays in reactor startup. The overall graphite oxidation in both plants was negligible

  16. Primary coolant chemistry of the Peach Bottom and Fort St. Vrain high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Burnette, R.D.; Baldwin, N.L.

    1981-01-01

    The chemical impurities in the primary coolants of the Peach Bottom and Fort St. Vrain reactors are discussed. The impurity mixtures in the two plants were quite different because the sources of the impurities were different. In the Peach Bottom reactor, the impurities were dominated by H 2 and CH 4 , which are decomposition products of oil. In the Fort St. Vrain reactor, there were high levels of CO, CO 2 , and H 2 O. Although oil ingress at Peach Bottom created carbon deposits on virtually all surfaces, its effect on reactor operation was negligible. Slow outgassing of water from the thermal insulation at Fort St. Vrain caused delays in reactor startup. The overall graphite oxidation in both plants was negligible. (author)

  17. A model for the description of the coolant mixing and its application to the analysis of boron dilution transients in pressurized water reactors; Ein Modell zur Beschreibung der Kuehlmittelvermischung und seine Anwendung auf die Analyse von Borverduennungstransienten in Druckwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren

    2010-08-15

    A model for the realistic description of the coolant mixing inside the pressure vessel of a pressurized water reactors has been developed and validated. This fast running model is based on the linear superposition of response functions on Dirac-pulse-like perturbations of the coolant parameters. It has been implemented into the coupled code system DYN3D/ATHLET nad serves as the interface between the one-dimensional thermal hydraulic system code ATHLETE and the 3D neutron kinetic core model DYN3D. By help of this model, the coolant mixing inside the reactor pressure vessel can e simulated in an efficient manner. A methodology for this analysis of hypothetical boron dilution accidents has been developed, which is based on the newly developed model for the coolant mixing. This methodology consists of a combination of stationary and transient calculations including a realistic treatment of the mixing of deborated slugs on the way towards the reactor core. The degree of conservatism can be adjusted by the variation of the initial size of the deborated slug. This new method was applied to two different boron dilution accidents. Besides the start of the first main coolant pump with a deborated slug of coolant in the cold leg of the primary circuit, a deboration event during the operation of the residual heat removal system was investigated. The results of the parameter study for a reactor core with a generic loading pattern demonstrated in both cases, that although the shut-down reactor becomes re-critical safety relevant cladding temperature limits are not reached, even if maximum possible volumes of the deborated slug are considered. The main reason for these results is the use of realistic time-dependent distributions of the boron concentration at the inlet of each fuel assembly.

  18. High pressure effects on fruits and vegetables

    OpenAIRE

    Timmermans, R.A.H.; Matser, A.M.

    2016-01-01

    The chapter provides an overview on different high pressure based treatments (high pressure pasteurization, blanching, pressure-assisted thermal processing, pressure-shift freezing and thawing) available for the preservation of fruits and vegetable products and extending their shelf life. Pressure treatment can be used for product modification through pressure gelatinization of starch and pressure denaturation of proteins. Key pressure–thermal treatment effects on vitamin, enzymes, flavor, co...

  19. Intraocular Pressure Increases After Intraarticular Knee Injection With Triamcinolone but Not Hyaluronic Acid.

    Science.gov (United States)

    Taliaferro, Kevin; Crawford, Alexander; Jabara, Justin; Lynch, Jonathan; Jung, Edward; Zvirbulis, Raimonds; Banka, Trevor

    2018-03-09

    Intraarticular steroid injections are a common first-line therapy for severe osteoarthritis, which affects an estimated 27 million people in the United States. Although topical, oral, intranasal, and inhalational steroids are known to increase intraocular pressure in some patients, the effect of intraarticular steroid injections on intraocular pressure has not been investigated, to the best of our knowledge. If elevated intraocular pressure is sustained for long periods of time or is of sufficient magnitude acutely, permanent loss of the visual field can occur. How does intraocular pressure change 1 week after an intraarticular knee injection either with triamcinolone acetonide or hyaluronic acid? A nonrandomized, nonblinded prospective cohort study was conducted at an outpatient, ambulatory orthopaedic clinic. This study compared intraocular pressure elevation before and 1 week after intraarticular knee injection of triamcinolone acetonide versus hyaluronic acid for management of primary osteoarthritis of the knee. Patients self-selected to be injected in their knee with either triamcinolone acetonide or hyaluronic acid before being informed of the study. The primary endpoint was intraocular pressure elevation of ≥ 7 mm Hg 1 week after injection. This cutoff is determined as the minimum significant pressure change in the ophthalmology literature recognized as an intermediate responder to steroids. Intraocular pressure was measured using a handheld Tono-Pen® applanation device. This device is frequently used in intraocular pressure measurement in clinical and research settings; 10 sequential measurements are obtained and averaged with a confidence interval. Only measurements with a 95% confidence interval were used. Over a 6-month period, a total of 96 patients were approached to enroll in the study. Sixty-two patients out of 96 approached (65%) agreed. Thirty-one (50%) were injected with triamcinolone and 31 (50%) were injected with hyaluronic acid. Patients

  20. Brillouin scattering at high pressures

    International Nuclear Information System (INIS)

    Grimsditch, M.; Polian, A.

    1988-02-01

    Technical advances which have made Brillouin scattering a useful tool in high pressure diamond anvil cell (DAC) studies, viz. multipassing and tandem operation of Fabry-Perot interferometers, are reviewed. Experimental aspects, such as allowed scattering geometries, are outlined and the data analysis required to transform Brillouin spectra into sound velocities and elastic constants is presented. Experimental results on H 2 , N 2 , Ar, and He are presented, and the close relationship between the Brillouin scattering results and equations of state is highlighted

  1. High pressure effects on fruits and vegetables

    NARCIS (Netherlands)

    Timmermans, R.A.H.; Matser, A.M.

    2016-01-01

    The chapter provides an overview on different high pressure based treatments (high pressure pasteurization, blanching, pressure-assisted thermal processing, pressure-shift freezing and thawing) available for the preservation of fruits and vegetable products and extending their shelf life. Pressure

  2. Management of large scale coolant channel replacement programme for Indian PHWRs

    International Nuclear Information System (INIS)

    Bhatnagar, V.K.; Chadda, S.K.; Arya, R.C.

    1994-01-01

    Coolant channel assemblies form most important core components of pressurised heavy water reactors. Zirconium alloy pressure tube which form part of coolant channel assemblies are subjected to environment of high neutron flux, high pressure and temperature. Under those operating environmental conditions, the pressure tubes material undergoes degradation of metallurgical and mechanical properties in addition to dimensional changes. The coolant channels are subjected to an in-service inspection (ISI) programme for monitoring the health particularly of the pressure tubes. The en-mass replacement of pressure tubes is needed after most of the pressure tubes show unacceptable conditions for an assured safe and reliable operation. An overview of various issues pertaining to this aspect is presented. (author). 4 figs

  3. Model Study of the Pressure Build-Up during Subcutaneous Injection

    DEFF Research Database (Denmark)

    Thomsen, Maria; Hernandez Garcia, Anier; Mathiesen, Joachim

    2014-01-01

    In this study we estimate the subcutaneous tissue counter pressure during drug infusion from a series of injections of insulin in type 2 diabetic patients using a non-invasive method. We construct a model for the pressure evolution in subcutaneous tissue based on mass continuity and the flow laws...

  4. Effect of high-temperature filtration on impurity composition in the primary circuit coolant of power units with WWER-1000 reactors

    International Nuclear Information System (INIS)

    Efimov, A.A.; Moskvin, L.N.; Gusev, B.A.; Leont'ev, G.G.; Nekrest'yanov, S.N.

    1992-01-01

    The effects of high-temperature filtration on changes in dispersive, chemical, radioisotope and phase compositions of impurities in primary circuit coolant of NPP with the WWER-1000 reactor are studied. Special filters are used for the studies. The data obtained confirm the applicability of high-temperature filtration for purification of WWER reactor water and steam separators at NPPs with RBMK reactors

  5. Atucha I nuclear power plant: Probabilistic safety study. Loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Perez, S.S.

    1987-01-01

    The plant response to the group of events 'large coolant loss' in order to evaluate the associated risk is analyzed. The event that covers all events of similar sequence due to its evolution features, being also the most demanded, is selected as starting event. The representative event is the 'guillotine type rupture of cold primary branch'. An annual occurrence frequency of 10/year is assumed for this event. The safety systems, when the event occurs, must assure the reactor shutdown and the core cooling, creating a heat sink to remove the decay heat. The annual frequency of core meltdown due to great loss of coolant is obtained multiplying the annual frequency of the starting event by the probability of failure of involved safety systems. By means of failure trees, the following is obtained: a) probability of failure to demand of the boron injection shutdown system = 4 x 10 -2 ; b) probability of failure to demand of the high pressure safety injection = 3 x 10 -3 ; c) probability of emergency cooling system failure = 4.4 x 10 -2 . Therefore, the three possible sequences of core meltdown have the following frequencies: λ 1 = 4 x 10 -6 /year λ 2 = 3 x 10 -7 /year λ 3 = 4.4 x 10 -6 /year. (Author)

  6. A simulation experiment and analysis on the effects of in-coherence in fuel coolant interaction

    International Nuclear Information System (INIS)

    Kondo, S.; Togo, Y.; Iwamura, T.

    1976-01-01

    Experimental and analytical studies were conducted to investigate effects of incoherence (space time behavior of molten fuel) on molten fuel coolant interaction. In experiments, a 2 mm diameter molten tin jet was injected upward into the water in a slender tank. The results were analyzed based on the pressure records and high speed photographs. The pressure records indicated that there were two types of interaction between molten jet and water, intermittent explosion mode and continuous one. The explosion mode appeared when the temperature of molten tin was above 350 0 C or so and that of water was below 70 0 C or so. The high speed photograph indicated that an establishment of a stable jet column was necessary for an explosive interaction and that a bubble like region grew and collapsed at the root of the jet in accordance with the generation of pressure pulse. It was found that the mass of metal which contributed to the vapor explosion was only a small part of the injected metal in the case of jet injection type contact mode and this was the reason why the gross thermal to mechanical energy conversion ratio was around 0.03% in this type of contact mode, though this ratio was around 2% if only the part of record around the pressure pulse was taken into consideration. In the analysis part, a multi-channel FCI model was developed to evaluate the spatial incoherence effect on pressure at subassembly exit. The calculated pressure trace indicated that the spatial incoherence has considerable effects for an evaluation of structure response under FCI pressure loading. (auth.)

  7. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  8. RELAP simulation and experimental verification of transient boiling conditions in narrow coolant channels, at low temperature and pressure

    International Nuclear Information System (INIS)

    Kunze, J.F.; Loyalka, S.K.; Hultsch, R.A.; Oladiran, O.; McKibben, J.C.

    1990-01-01

    This paper reports on benchmark experiments needed to verify the accuracy of thermal hydraulic codes (such as RELAP5/MOD2) with respect to their capability to simulate transient boiling conditions both with and without a closed recirculation path in narrow channels, under essentially atmospheric pressure conditions characteristic of plate-type research reactors. An experimental apparatus with this objective has been constructed, and data for surface heat flux of 1.2 x 10 5 w/m 2 are reported

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

    International Nuclear Information System (INIS)

    Schatz, K.

    1982-01-01

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

  10. A Comparative Study on Energy and Exergy Analyses of a CI Engine Performed with Different Multiple Injection Strategies at Part Load: Effect of Injection Pressure

    Directory of Open Access Journals (Sweden)

    Muammer Özkan

    2015-01-01

    Full Text Available In this study, a four stroke four cylinder direct injection CI engine was run using three different injection pressures. In all measurements, the fuel quantity per cycle, the pre injection and main injection timing, the boost pressure and the engine speed were kept constant. The motor tests were performed under 130, 140 and 150 MPa rail pressure. During the theoretical part of the study, combustion, emission, energy and exergy analysis were made using the test results. An increase in the injection pressure increases combustion efficiency. The results show that combustion efficiency is not enough by itself, because the increase in the power need of the injection pump, decreases the thermal efficiency. The increase in the combustion temperature, increases the cooling loss and decreases the exergetic efficiency. In addition, the NOx emissions increased by 12% and soot emissions decreased 44% via increasing injection pressure by 17%. The thermal and exergetic efficiencies are found inversely proportional with injection pressure. Exergy destruction is found independent of the injection pressure and its value is obtained as ~6%.

  11. Adsorption purification of helium coolant of high-temperature gas-cooled reactors of carbon dioxide

    International Nuclear Information System (INIS)

    Varezhkin, A.V.; Zel'venskij, Ya.D.; Metlik, I.V.; Khrulev, A.A.; Fedoseenkin, A.N.

    1986-01-01

    A series experiments on adsorption purification of helium of CO 2 using national adsorbent under the conditions characteristic of HTGR type reactors cleanup system is performed. The experimnts have been conducted under the dynamic mode with immobile adsorbent layer (CaA zeolite) at gas flow rates from 0,02 to 0,055 m/s in the pressure range from 0,8 to 5 MPa at the temperature of 273 and 293 K. It is shown that the adsorption grows with the decrease of gas rate, i.e. with increase of contact time with adsorbent. The helium pressure, growth noticeably whereas the temperature decrease from 293 to 273 K results in adsorption 2,6 times increase. The conclusion is drawn that it is advisable drying and purification of helium of CO 2 to perform separately using different zeolites: NaA - for water. CaA - for CO 2 . Estimations of purification unit parameters are realized

  12. High Pressure Electrolyzer System Evaluation

    Science.gov (United States)

    Prokopius, Kevin; Coloza, Anthony

    2010-01-01

    This report documents the continuing efforts to evaluate the operational state of a high pressure PEM based electrolyzer located at the NASA Glenn Research Center. This electrolyzer is a prototype system built by General Electric and refurbished by Hamilton Standard (now named Hamilton Sunstrand). It is capable of producing hydrogen and oxygen at an output pressure of 3000 psi. The electrolyzer has been in storage for a number of years. Evaluation and testing was performed to determine the state of the electrolyzer and provide an estimate of the cost for refurbishment. Pressure testing was performed using nitrogen gas through the oxygen ports to ascertain the status of the internal membranes and seals. It was determined that the integrity of the electrolyzer stack was good as there were no appreciable leaks in the membranes or seals within the stack. In addition to the integrity testing, an itemized list and part cost estimate was produced for the components of the electrolyzer system. An evaluation of the system s present state and an estimate of the cost to bring it back to operational status was also produced.

  13. Welding lines formation in holes obtained by low pressure injection molding of ceramic parts

    Directory of Open Access Journals (Sweden)

    C. A. Costa

    Full Text Available Abstract This work presents a study to evaluate the process of producing internal holes in ceramic disks produced by low pressure injection molding (LPIM process. Two process conditions defined as pre-injection and post-injection were used to test the proposition. In the first one the pin cores that produce the holes were positioned in the cavity before the injection of the feedstock; and in the second one, the pin cores were positioned in the cavity, just after the feeding phase of the injection mold. An experimental injection mold designed and manufactured to test both processes was developed to produce ceramic disk with Ø 50 x 2 mm with four holes of Ø 5 mm, equally and radially distributed through the disk. The feedstock was composed of 86 wt% alumina (Al2O3 and 14 wt% organic vehicle based on paraffin wax. Heating and cooling systems controlled by a data acquisition system were included in the mold. The results showed that there were no welding lines with the post-injection process, proving to be an option for creating holes in the ceramic parts produced by LPIM. It was observed that best results were obtained at 58 °C mold temperature. The pins extraction temperature was about 45 °C, and the injection pressure was 170 kPa.

  14. Effect of fuel injection pressure and injection timing of Karanja biodiesel blends on fuel spray, engine performance, emissions and combustion characteristics

    International Nuclear Information System (INIS)

    Agarwal, Avinash Kumar; Dhar, Atul; Gupta, Jai Gopal; Kim, Woong Il; Choi, Kibong; Lee, Chang Sik; Park, Sungwook

    2015-01-01

    Highlights: • Effect of FIP on microscopic spray characteristics. • Effect of FIP and SOI timing on CRDI engine performance, emissions and combustion. • Fuel injection duration shortened, peak injection rate increased with increasing FIP. • SMD (D 32 ) and AMD (D 10 ) of fuel droplets decreased for lower biodiesel blends. • Increase in biodiesel blend ratio and FIP, fuel injection duration decreased. - Abstract: In this investigation, effect of 10%, 20% and 50% Karanja biodiesel blends on injection rate, atomization, engine performance, emissions and combustion characteristics of common rail direct injection (CRDI) type fuel injection system were evaluated in a single cylinder research engine at 300, 500, 750 and 1000 bar fuel injection pressures at different start of injection timings and constant engine speed of 1500 rpm. The duration of fuel injection slightly decreased with increasing blend ratio of biodiesel (Karanja Oil Methyl Ester: KOME) and significantly decreased with increasing fuel injection pressure. The injection rate profile and Sauter mean diameter (D 32 ) of the fuel droplets are influenced by the injection pressure. Increasing fuel injection pressure generally improves the thermal efficiency of the test fuels. Sauter mean diameter (D 32 ) and arithmetic mean diameter (D 10 ) decreased with decreasing Karanja biodiesel content in the blend and significantly increased for higher blends due to relatively higher fuel density and viscosity. Maximum thermal efficiency was observed at the same injection timing for biodiesel blends and mineral diesel. Lower Karanja biodiesel blends (up to 20%) showed lower brake specific hydrocarbon (BSHC) and carbon monoxide (BSCO) emissions in comparison to mineral diesel. For lower Karanja biodiesel blends, combustion duration was shorter than mineral diesel however at higher fuel injection pressures, combustion duration of 50% blend was longer than mineral diesel. Up to 10% Karanja biodiesel blends in a CRDI

  15. Experimental investigation of the concomitant injection of gasoline and CNG in a turbocharged spark ignition engine

    International Nuclear Information System (INIS)

    Momeni Movahed, M.; Basirat Tabrizi, H.; Mirsalim, M.

    2014-01-01

    Highlights: • Concomitant injection of gasoline and CNG is compared with gasoline and CNG modes. • BSFC, HC and CO emissions of the concomitant injection are lower than gasoline mode. • Deteriorations of the concomitant injection are negligible compared to gasoline mode. • Cylinder peak pressure and heat loss to coolant of the concomitant injection are lower than CNG mode. • Some shortcomings in CNG mode can be solved by changing the spark timing and lambda. - Abstract: Concomitant injection of gasoline and CNG is a new concept to overcome problems of bi-fueled spark ignition engines, which operate in single fuel mode, either in gasoline or in CNG mode. This experimental study indicates how some problems of gasoline mode such as retarded ignition timings for knock prevention and rich air–fuel mixture for component protection can be resolved with the concomitant injection of gasoline and CNG. Results clearly show that the concomitant injection improves thermal efficiency compared to gasoline mode. On the other hand, simultaneous injection of gasoline and CNG reduces some problems of CNG mode such as high cylinder pressure and heat loss to the engine coolant. This decreases the stringent requirements for thermal and mechanical strength of the engine components in CNG mode. In addition, it is shown that by modifying the spark advance and air fuel ratio in CNG mode, the engine operation improves in terms of NOx emissions and maximum in-cylinder pressure as the concomitant injection does. Nevertheless, new requirements such as an intercooler with higher cooling capacity are implied to the engine configuration. Finally, the most important concerns in control strategies of the engine control unit for a vehicle with concomitant injection of gasoline and CNG are discussed

  16. Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

    International Nuclear Information System (INIS)

    Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

    2000-01-01

    The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

  17. Development of a deformation and failure model for Zircaloy at high temperatures for light water reactor loss-of-coolant-accident investigations

    International Nuclear Information System (INIS)

    Raff, S.

    1982-11-01

    To describe Zircaloy-4 deformation and failure behaviour at high temperatures (600 to 1400 0 C), the phenomenological model NORA was developed and verified against numerous experimental results. The model can be applied to the calculation of fuel rod cladding deformation during small and large break loss-of-coolant-accidents. (orig./RW) [de

  18. Effect of injection pressure on heat release rate and emissions in CI engine using orange skin powder diesel solution

    International Nuclear Information System (INIS)

    Purushothaman, K.; Nagarajan, G.

    2009-01-01

    Experiments have been conducted to study the effect of injection pressure on the combustion process and exhaust emissions of a direct injection diesel engine fueled with Orange Skin Powder Diesel Solution (OSPDS). Earlier investigation by the authors revealed that 30% OSPDS was optimum for better performance and emissions. In the present investigation the injection pressure was varied with 30% OSPDS and the combustion, performance and emissions characteristics were compared with those of diesel fuel. The different injection pressures studied were 215 bar, 235 bar and 255 bar. The results showed that the cylinder pressure with 30% OSPDS at 235 bar fuel injection pressure, was higher than that of diesel fuel as well as at other injection pressures. Similarly, the ignition delay was longer and with shorter combustion duration with 30% OSPDS at 235 bar injection pressure. The brake thermal efficiency was better at 235 bar than that of other fuel injection pressures with OSPDS and lower than that of diesel fuel. The NO x emission with 30% OSPDS was higher at 235 bar. The hydrocarbon and CO emissions were lower with 30% OSPDS at 235 bar. The smoke emission with 30% OSPDS was marginally lower at 235 bar and marginally higher at 215 bar than for diesel fuel. The combustion, performance and emission characteristics of the engine operating on the test fuels at 235 bar injection pressure were better than other injection pressures

  19. Stresses and strains in the steel containment resulting from transient pressure and temperature loading during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gruner, P.; Kuntze, W.M.; Jansky, J.

    1985-01-01

    Posttest calculations of stresses and strains in the steel containment of the German research reactor HDR were performed for a simulated LOCA. The results of the theoretical investigations are presented and compared to experimental findings. The pressure and temperature loading of the shell was determined with the thermodynamic code COFLOW on the basis of a multi-compartment model. Using a three-dimensional finite element model the temporal behaviour of the containment was calculated employing the structural mechanics code ASKA. Global bending deformations and local negative straining of the steel shell is discussed. Theoretical and experimental results agree in most cases rather well. Reasons for deviations will be discussed. The specific behaviour of strains found in the vicinity of locally heated areas will be explained by means of analytical considerations. (orig.)

  20. Pilot test of pressure maintenance by water injection and gas injection in the M'Bega field

    Energy Technology Data Exchange (ETDEWEB)

    Hodee, B

    1965-02-01

    The M'Bega reservoir, Gabon, is one that is both fractured and fissured, and its reservoir rock, improperly referred to as silicified clay, is made up of opal-cement silt. At the time it began production, the major unknown factor in this field was the amount of oil it contained. The factor was finally determined after 18 mo. of production, once data was obtained concerning the advance of aquifiers. Up until that time, applied comparisons of successive data could not differentiate between the activity of aquifers and that of dissolved gas expansion with partial segregation. Consequently, a pilot test was made in which pressure was maintained by water injection. Then, with a double drainage phenomena (the tops by the gas, and the flanks by the aquifer water), production of the field consisted of bringing about a coincidence between the water and gas fronts at the level of the existing wells by means of gas injection.

  1. Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

    1980-01-01

    A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

  2. Molding Properties of Inconel 718 Feedstocks Used in Low-Pressure Powder Injection Molding

    Directory of Open Access Journals (Sweden)

    Fouad Fareh

    2016-01-01

    Full Text Available The impact of binders and temperature on the rheological properties of feedstocks used in low-pressure powder injection molding was investigated. Experiments were conducted on different feedstock formulations obtained by mixing Inconel 718 powder with wax-based binder systems. The shear rate sensitivity index and the activation energy were used to study the degree of dependence of shear rate and temperature on the viscosity of the feedstocks. The injection performance of feedstocks was then evaluated using an analytical moldability model. The results indicated that the viscosity profiles of feedstocks depend significantly on the binder constituents, and the secondary binder constituents play an important role in the rheological behavior (pseudoplastic or near-Newtonian exhibited by the feedstock formulations. Viscosity values as low as 0.06 to 2.9 Pa·s were measured at high shear rates and high temperatures. The results indicate that a feedstock containing a surfactant agent exhibits the best moldability characteristics.

  3. Infusion pressure and pain during microneedle injection into skin of human subjects

    Science.gov (United States)

    Gupta, Jyoti; Park, Sohyun; Bondy, Brian; Felner, Eric I.; Prausnitz, Mark R.

    2011-01-01

    Infusion into skin using hollow microneedles offers an attractive alternative to hypodermic needle injections. However, the fluid mechanics and pain associated with injection into skin using a microneedle have not been studied in detail before. Here, we report on the effect of microneedle insertion depth into skin, partial needle retraction, fluid infusion flow rate and the co-administration of hyaluronidase on infusion pressure during microneedle-based saline infusion, as well as on associated pain in human subjects. Infusion of up to a few hundred microliters of fluid required pressures of a few hundred mmHg, caused little to no pain, and showed weak dependence on infusion parameters. Infusion of larger volumes up to 1 mL required pressures up to a few thousand mmHg, but still usually caused little pain. In general, injection of larger volumes of fluid required larger pressures and application of larger pressures cause more pain, although other experimental parameters also played a significant role. Among the intradermal microneedle groups, microneedle length had little effect; microneedle retraction lowered infusion pressure but increased pain; lower flow rate reduced infusion pressure and kept pain low; and use of hyaluronidase also lowered infusion pressure and kept pain low. We conclude that microneedles offer a simple method to infuse fluid into the skin that can be carried out with little to no pain. PMID:21684001

  4. High pressure research at CHESS

    International Nuclear Information System (INIS)

    Brister, K.

    1992-01-01

    Since February 1990 there has been a dedicated high pressure line at the Cornell High Energy Synchrotron Source (CHESS). This facility provides X-ray instrumentation for energy dispersive X-ray diffraction and Laue diffraction using diamond anvil cells. Both hard-bend magnet and wiggler radiation are available as well as focused monochromatic radiation. In addition, support instrumentation is also available; a ruby system, laser heating, sample loading, and data analysis software. Experienced users need only to bring their diamond anvil cells and samples and can leave with the initial data analysis finished. Research using diamond anvil cells will be introduced and the facility will be described. Some of the diamond anvil cell research done at CHESS will be reviewed, including crystalline to amorphous transitions (R.R. Winters et al., Chem. Phys, in press), properties of C 6 0 under stress (S.J. Duclos et al., Nature 351 (1991) 380), deep earthquakes (T.C. Wu et al., submitted to J. Geophys. Res.)l, and reaching pressures of the center of Earth (A.L. Ruoff et al., Rev. Sci. Instr. 61 (1990) 3830). (orig.)

  5. Studies on Microscopic Structure of Diesel Sprays under Atmospheric and High Gas Pressures

    Directory of Open Access Journals (Sweden)

    D. Deshmukh

    2014-06-01

    Full Text Available In the present work, the spray structure of diesel from a 200-μm, single-hole solenoid injector is studied using microscopic imaging at injection pressures of 700, 1000 and 1400 bar for various gas pressures. A long-distance microscope with a high resolution camera is used for spray visualization with a direct imaging technique. This study shows that even at very high injection pressures, the spray structure in an ambient environment of atmospheric pressure reveals presence of entangled ligaments and non-spherical droplets during the injection period. With increase in the injection pressure, the ligaments tend to get smaller and spread radially. The spray structure studies are also conducted at high gas pressures in a specially designed high pressure chamber with optical access. The near nozzle spray structure at the end of the injection shows that the liquid jet breakup is improved with increase in gas density. The droplet size measurement is possible only late in the injection duration when the breakup appears to be complete and mostly spherical droplets are observed. Hence, droplet size measurements are performed after 1.3 ms from start of the injection pulse. Spatial and temporal variation in Sauter Mean Diameter (SMD is observed and reported for the case corresponding to an injection pressure of 700 bar. Overall, this study has highlighted the importance of verifying the extentof atomization and droplet shape even in dense sprays before using conventional dropsizing methods such as PDPA.

  6. Studies on dissolution characteristics of simulated corrosion products on pressurized water reactor primary coolant loops. Pt.2: Cobalt simulated corrosion product

    International Nuclear Information System (INIS)

    Li Shan; Zhou Xianyu

    1997-01-01

    The studies on the dissolution characteristics of simulated corrosion product of cobalt on pressurized water reactor primary coolant loops in aqueous solution of citric acid, hydrogen peroxide and citric acid-hydrogen peroxide have been performed. The results show that the portion of the dissolved simulated corrosion product of cobalt in citric acid aqueous solution clearly increases with a rise in citric acid concentration and is ten times above the corresponding value of iron. The portion of the products that dissolve is the largest at pH 3.00 in the pH range of 2.33∼4.50 and at 70 degree C in the range of 60∼80 degree C. It is shown that the portion of the dissolved simulated corrosion product of cobalt in hydrogen peroxide aqueous solution is smaller than the corresponding value in citric acid, and that the portion of the dissolved simulated corrosion product of cobalt in aqueous solution of hydrogen peroxide-citric acid is larger than the corresponding value in single citric acid aqueous solution

  7. Evaluation of the radiative transfer in the core of a Pressurized Water Reactor (PWR) during the reflooding step of a Loss Of Coolant Accident (LOCA)

    International Nuclear Information System (INIS)

    Gerardin, J.

    2012-01-01

    We developed a method of resolution of radiative transfer inside a medium of vapor-droplets surrounded by hot walls, in order to couple it with a simulation of the flow at the CFD scale. The scope is the study of the cooling of the core of nuclear reactor following a Loss Of Coolant Accident (LOCA). The problem of radiative transfer can be cut into two sub problems, one concerning the evaluation of the radiative properties of the medium and a second concerning the solution of the radiative transfer equation. The radiative properties of the droplets have been computed with the use of the Mie Theory and those of the vapor have been computed with a Ck model. The medium made of vapor and droplets is an absorbing, anisotropically scattering, emissive, non grey, non homogeneous medium. Hence, owing to the possible variations of the flow properties (diameter and volumetric fraction of the droplets, temperature and pressure of the vapor), the medium can be optically thin or thick. Consequently, a method is required which solves the radiative transfer accurately, with a moderate calculation time for all of these prerequisites. The IDA has been chosen, derived from the well-known P1-approximation. Its accuracy has been checked on academical cases found in the literature and by comparison with experimental data. Simulations of LOCA flows have been conducted taking account of the radiative transfer, evaluating the radiative fluxes and showing that radiative transfer influence cannot be neglected. (author)

  8. Analysis of Cavity Pressure and Warpage of Polyoxymethylene Thin Walled Injection Molded Parts: Experiments and Simulations

    DEFF Research Database (Denmark)

    Guerrier, Patrick; Tosello, Guido; Hattel, Jesper Henri

    2014-01-01

    Process analysis and simulations on molding experiments of 3D thin shell parts have been conducted. Moldings were carried out with polyoxymethylene (POM). The moldings were performed with cavity pressure sensors in order to compare experimental process results with simulations. The warpage...... was characterized by measuring distances using a tactile coordinate measuring machine (CMM). Molding simulations have been executed taking into account actual processing conditions. Various aspects have been considered in the simulation: machine barrel geometry, injection speed profiles, cavity injection pressure......, melt and mold temperatures, material rheological and pvT characterization. Factors investigated for comparisons were: injection pressure profile, short shots length, flow pattern, and warpage. A reliable molding experimental database was obtained, accurate simulations were conducted and a number...

  9. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  10. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  11. Development of micropump-actuated negative pressure pinched injection for parallel electrophoresis on array microfluidic chip.

    Science.gov (United States)

    Li, Bowei; Jiang, Lei; Xie, Hua; Gao, Yan; Qin, Jianhua; Lin, Bingcheng

    2009-09-01

    A micropump-actuated negative pressure pinched injection method is developed for parallel electrophoresis on a multi-channel LIF detection system. The system has a home-made device that could individually control 16-port solenoid valves and a high-voltage power supply. The laser beam is excitated and distributes to the array separation channels for detection. The hybrid Glass-PDMS microfluidic chip comprises two common reservoirs, four separation channels coupled to their respective pneumatic micropumps and two reference channels. Due to use of pressure as a driving force, the proposed method has no sample bias effect for separation. There is only one high-voltage supply needed for separation without relying on the number of channels, which is significant for high-throughput analysis, and the time for sample loading is shortened to 1 s. In addition, the integrated micropumps can provide the versatile interface for coupling with other function units to satisfy the complicated demands. The performance is verified by separation of DNA marker and Hepatitis B virus DNA samples. And this method is also expected to show the potential throughput for the DNA analysis in the field of disease diagnosis.

  12. Continuous positive airway pressure breathing increases cranial spread of sensory blockade after cervicothoracic epidural injection of lidocaine.

    NARCIS (Netherlands)

    Visser, W.A.; Eerd, M.J. van; Seventer, R. van; Gielen, M.J.M.; Giele, J.L.P.; Scheffer, G.J.

    2007-01-01

    BACKGROUND: Continuous positive airway pressure (CPAP) increases the caudad spread of sensory blockade after low-thoracic epidural injection of lidocaine. We hypothesized that CPAP would increase cephalad spread of blockade after cervicothoracic epidural injection. METHODS: Twenty patients with an

  13. High-pressure microhydraulic actuator

    Science.gov (United States)

    Mosier, Bruce P [San Francisco, CA; Crocker, Robert W [Fremont, CA; Patel, Kamlesh D [Dublin, CA

    2008-06-10

    Electrokinetic ("EK") pumps convert electric to mechanical work when an electric field exerts a body force on ions in the Debye layer of a fluid in a packed bed, which then viscously drags the fluid. Porous silica and polymer monoliths (2.5-mm O.D., and 6-mm to 10-mm length) having a narrow pore size distribution have been developed that are capable of large pressure gradients (250-500 psi/mm) when large electric fields (1000-1500 V/cm) are applied. Flowrates up to 200 .mu.L/min and delivery pressures up to 1200 psi have been demonstrated. Forces up to 5 lb-force at 0.5 mm/s (12 mW) have been demonstrated with a battery-powered DC-DC converter. Hydraulic power of 17 mW (900 psi@ 180 uL/min) has been demonstrated with wall-powered high voltage supplies. The force and stroke delivered by an actuator utilizing an EK pump are shown to exceed the output of solenoids, stepper motors, and DC motors of similar size, despite the low thermodynamic efficiency.

  14. Nuclear magnetic resonance studies at high pressures

    International Nuclear Information System (INIS)

    Jonas, J.

    1980-01-01

    Recent advances in the field of NMR spectroscopy at high pressure are reviewed. After a brief discussion of two novel experimental techniques, the main focus of this review is on several specific studies which illustrate the versatility and power of this high pressure field. Experimental aspects of NMR measurements at high pressure and high temperature and the techniques for the high resolution NMR spectroscopy at high pressure are discussed. An overview of NMR studies of the dynamic structure of simple polyatomic liquids and hydrogen bonded liquids is followed by a discussion of high resolution spectroscopy at high pressure. Examples of NMR studies of disordered organic solids and polymers conclude the review. (author)

  15. Congestion of mastoid mucosa and influence on middle ear pressure - Effect of retroauricular injection of adrenaline.

    Science.gov (United States)

    Fooken Jensen, Pernille Vita; Gaihede, Michael

    2016-10-01

    Micro-CT scanning of temporal bones has revealed numerous retroauricular microchannels, which connect the outer bone surface directly to the underlying mastoid air cells. Their structure and dimensions have suggested a separate vascular supply to the mastoid mucosa, which may play a role in middle ear (ME) pressure regulation. This role may be accomplished by changes in the mucosa congestion resulting in volumetric changes, which ultimately affect the pressure of the enclosed ME gas pocket (Boyle's law). Further, such mucosa congestion may be susceptible to α-adrenergic stimulation similar to the mucosa of the nose. The purpose of our study was to investigate these hypotheses by recording the ME pressure in response to adrenergic stimulation administered by retroauricular injections at the surface of the microchannels. In a group of 20 healthy adults we measured the ME pressure by tympanometry initially in the sitting position, and then in the supine position over a 5 min period with 30 s intervals. In each subject, the study included 1) a control reference experiment with no intervention, 2) a control experiment with subcutaneously retroauricular injection of 1 ml isotonic NaCl solution, and 3) a test experiment with subcutaneously retroauricular injection of 1 ml NaCl-adrenaline solution. In both control experiments the ME pressure displayed an immediate increase in response to changing body position; this pressure increase remained stable for the entire period up to five minutes. In the test experiments the ME pressure also showed an initial pressure increase, but it was followed by a distinct significant pressure decrease with a maximum after 90 s. The test group was injected with both a 5 and 10% adrenaline solution, but the responses appeared similar for the two concentrations. Subcutaneous retroauricular injection of adrenaline caused a significant pressure decrease in ME pressure compared with control ears. This may be explained by the microchannels

  16. DASH diet to lower high blood pressure

    Science.gov (United States)

    ... patientinstructions/000770.htm DASH diet to lower high blood pressure To use the sharing features on this page, ... Hypertension. The DASH diet can help lower high blood pressure and cholesterol and other fats in your blood. ...

  17. What Is High Blood Pressure Medicine?

    Science.gov (United States)

    ... a medicine calendar. • Set a reminder on your smartphone. What types of medicine may be prescribed? One ... High Blood Pressure Medicine? What are their side effects? For many people, high blood pressure medicine can ...

  18. High blood pressure - medicine-related

    Science.gov (United States)

    Drug-induced hypertension is high blood pressure caused by using a chemical substance or medicine. ... of the arteries There are several types of high blood pressure : Essential hypertension has no cause that can be ...

  19. High blood pressure and eye disease

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/000999.htm High blood pressure and eye disease To use the sharing features on this page, please enable JavaScript. High blood pressure can damage blood vessels in the retina . The ...

  20. Experiments on aerosol removal by high-pressure water spray

    Energy Technology Data Exchange (ETDEWEB)

    Corno, Ada del, E-mail: delcorno@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Morandi, Sonia, E-mail: morandi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Parozzi, Flavio, E-mail: parozzi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Araneo, Lucio, E-mail: lucio.araneo@polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy); CNR-IENI, via Cozzi 53, I-20125 Milano (Italy); Casella, Francesco, E-mail: francesco2.casella@mail.polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy)

    2017-01-15

    Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m{sup 3}. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m{sup 3}. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was

  1. Experiments on aerosol removal by high-pressure water spray

    International Nuclear Information System (INIS)

    Corno, Ada del; Morandi, Sonia; Parozzi, Flavio; Araneo, Lucio; Casella, Francesco

    2017-01-01

    Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m"3. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m"3. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was detected with 1

  2. High pressure metrology for industrial applications

    Science.gov (United States)

    Sabuga, Wladimir; Rabault, Thierry; Wüthrich, Christian; Pražák, Dominik; Chytil, Miroslav; Brouwer, Ludwig; Ahmed, Ahmed D. S.

    2017-12-01

    To meet the needs of industries using high pressure technologies, in traceable, reliable and accurate pressure measurements, a joint research project of the five national metrology institutes and the university was carried out within the European Metrology Research Programme. In particular, finite element methods were established for stress-strain analysis of elastic and nonlinear elastic-plastic deformation, as well as of contact processes in pressure-measuring piston-cylinder assemblies, and high-pressure components at pressures above 1 GPa. New pressure measuring multipliers were developed and characterised, which allow realisation of the pressure scale up to 1.6 GPa. This characterisation is based on research including measurements of material elastic constants by the resonant ultrasound spectroscopy, hardness of materials of high pressure components, density and viscosity of pressure transmitting liquids at pressures up to 1.4 GPa and dimensional measurements on piston-cylinders. A 1.6 GPa pressure system was created for operation of the 1.6 GPa multipliers and calibration of high pressure transducers. A transfer standard for 1.5 GPa pressure range, based on pressure transducers, was built and tested. Herewith, the project developed the capability of measuring pressures up to 1.6 GPa, from which industrial users can calibrate their pressure measurement devices for accurate measurements up to 1.5 GPa.

  3. Double Tunneling Injection Quantum Dot Lasers for High Speed Operation

    Science.gov (United States)

    2017-10-23

    Double Tunneling-Injection Quantum Dot Lasers for High -Speed Operation The views, opinions and/or findings contained in this report are those of...SECURITY CLASSIFICATION OF: 1. REPORT DATE (DD-MM-YYYY) 4. TITLE AND SUBTITLE 13. SUPPLEMENTARY NOTES 12. DISTRIBUTION AVAILIBILITY STATEMENT 6...State University Title: Double Tunneling-Injection Quantum Dot Lasers for High -Speed Operation Report Term: 0-Other Email: asryan@vt.edu Distribution

  4. Coolant leakage detection device

    International Nuclear Information System (INIS)

    Ito, Takao.

    1983-01-01

    Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

  5. Low-pressure injection molding of alumina ceramics using a carnauba wax binder: preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Quevedo Nogueira, R.E.F.; Bezerra, A.C.; Santos, F.C. dos [Dept. de Engenharia Mecanica, Centro de Tecnologia-UFC, Fortaleza, CE (Brazil); Sousa, M.R. de; Acchar, W. [Dept. de Engenharia Mecanica, Univ. Federal do Rio Grande do Norte, UFRN-Campus Univ., Natal, RN (Brazil)

    2001-07-01

    Carnauba wax, a natural product from Northeastern Brazil, has found application in the processing of ceramics. However, the use of pure carnauba wax is not recommended due to its narrow melting range and poor mechanical properties. In the present work carnauba wax based organic vehicles with the addition of low-density polyethylene and stearic acid were developed for use in the low-pressure injection molding of alumina ceramics. Viscosimetric testing was employed for the determination of optimal composition of the organic vehicle. The optimal content of ceramic powder in the mixture was also determined. All the materials used are easily available in the Brazilian market. A simple ceramic part was injected at low pressures (0.6 MPa) using a semi-automatic injection molding machine. For this purpose a double cavity mold was designed and built. Preliminary results demonstrate the technical viability of the process using the organic vehicle developed. (orig.)

  6. Effect of pegaptanib sodium 0.3 mg intravitreal injections (Macugen) in intraocular pressure

    DEFF Research Database (Denmark)

    Boyer, David S; Goldbaum, Mauro; Leys, Anita M

    2014-01-01

    OBJECTIVE: To assess the rate of pegaptanib-associated sustained intraocular pressure (IOP) elevation. METHODS: A posthoc analysis was conducted on all IOP measurements, except the immediate 30-min postinjection, from all subjects randomised to pegaptanib 0.3 mg or sham injections continuously in...

  7. Experimental in situ investigations of turbulence under high pressure.

    Science.gov (United States)

    Song, Kwonyul; Al-Salaymeh, Ahmed; Jovanovic, Jovan; Rauh, Cornelia; Delgado, Antonio

    2010-02-01

    In tube injection systems applied in high-pressure processing of packed biomaterials and foods, the pressure-transmitting medium is injected into the vessel to increase the pressure up to 1000 MPa, generating a submerged liquid-free jet. The presence of a turbulent-free jet during the pressurization phase and its positive influence on the homogeneity of the product treatment has already been examined by computational fluid dynamics investigations. However, no experimental data have supported the existence and properties of turbulent flow under high-pressure (HP) conditions up to 400 MPa. This contribution presents the development of two experimental setups: HP-laser Doppler anemometry and HP-hot wire anemometry. For the first time the time-averaged velocity profiles of a free jet during pressurization up to 300 MPa at different Reynolds numbers (Re) have been obtained. In this article, the dependence of the velocity profiles on the Re is discussed in detail. Moreover, the relaminarization phenomenon of the turbulent pipe flow most likely caused by the compressibility effects and viscosity changes of the pressure-transmitting medium is examined.

  8. High-Pressure Lightweight Thrusters

    Science.gov (United States)

    Holmes, Richard; McKechnie, Timothy; Shchetkovskiy, Anatoliy; Smirnov, Alexander

    2013-01-01

    Returning samples of Martian soil and rock to Earth is of great interest to scientists. There were numerous studies to evaluate Mars Sample Return (MSR) mission architectures, technology needs, development plans, and requirements. The largest propulsion risk element of the MSR mission is the Mars Ascent Vehicle (MAV). Along with the baseline solid-propellant vehicle, liquid propellants have been considered. Similar requirements apply to other lander ascent engines and reaction control systems. The performance of current state-ofthe- art liquid propellant engines can be significantly improved by increasing both combustion temperature and pressure. Pump-fed propulsion is suggested for a single-stage bipropellant MAV. Achieving a 90-percent stage propellant fraction is thought to be possible on a 100-kg scale, including sufficient thrust for lifting off Mars. To increase the performance of storable bipropellant rocket engines, a high-pressure, lightweight combustion chamber was designed. Iridium liner electrodeposition was investigated on complex-shaped thrust chamber mandrels. Dense, uniform iridium liners were produced on chamber and cylindrical mandrels. Carbon/carbon composite (C/C) structures were braided over iridium-lined mandrels and densified by chemical vapor infiltration. Niobium deposition was evaluated for forming a metallic attachment flange on the carbon/ carbon structure. The new thrust chamber was designed to exceed state-of-the-art performance, and was manufactured with an 83-percent weight savings. High-performance C/Cs possess a unique set of properties that make them desirable materials for high-temperature structures used in rocket propulsion components, hypersonic vehicles, and aircraft brakes. In particular, more attention is focused on 3D braided C/Cs due to their mesh-work structure. Research on the properties of C/Cs has shown that the strength of composites is strongly affected by the fiber-matrix interfacial bonding, and that weakening

  9. Evaluation of high temperature pressure sensors

    International Nuclear Information System (INIS)

    Choi, In-Mook; Woo, Sam-Yong; Kim, Yong-Kyu

    2011-01-01

    It is becoming more important to measure the pressure in high temperature environments in many industrial fields. However, there is no appropriate evaluation system and compensation method for high temperature pressure sensors since most pressure standards have been established at room temperature. In order to evaluate the high temperature pressure sensors used in harsh environments, such as high temperatures above 250 deg. C, a specialized system has been constructed and evaluated in this study. The pressure standard established at room temperature is connected to a high temperature pressure sensor through a chiller. The sensor can be evaluated in conditions of changing standard pressures at constant temperatures and of changing temperatures at constant pressures. According to the evaluation conditions, two compensation methods are proposed to eliminate deviation due to sensitivity changes and nonlinear behaviors except thermal hysteresis.

  10. Method of decontaminating primary coolant circuits

    International Nuclear Information System (INIS)

    Ishibashi, Masaru; Sumi, Masao.

    1981-01-01

    Purpose: To eliminate hard contaminated layers as well as soft contaminated layers without injuring substrate materials, upon decontamination of radiation contaminated portions in equipments and pipeways constituting primary coolant circuits. Constitution: High pressure water from a high pressure pump is jetted out from the nozzle of a spray gun to the radiation contaminated portions in equipments, for example, to the surface of water chamber in a vapor evaporator. High pressure pure water or aqueous boric acid is jetted out from the periphery and boric oxide particles (of about 1 - 100 μ particle size) are jetted out from the center of the nozzle of the spray gun. The particles (blasting material) jetted out together with the high pressure water impinge on the contaminated surfaces to remove the contaminated layers. Upon impingement, the high pressure water acts as the shock absorber for the blasting material and, after the impingement, it flows down to the bottom of the water chamber, and the blasting material is dissolved in the high pressure water. (Horiuchi, T.)

  11. 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

  12. Influence of the pressure holding time on strain generation in fuel injection lines

    International Nuclear Information System (INIS)

    Basara, Adis; Alt, Nicolas; Schluecker, Eberhard

    2011-01-01

    An influence of the pressure holding time on residual strain generation during the autofrettage process was studied experimentally for the first time in the present work. It is the state of the art that fuel injection lines are held at the autofrettage pressure for only a few seconds in an industrial production. In doing so, it is assumed that a desirable residual stress-strain pattern is generated. However, the results of the experimental investigations outlined in this work indicated that completion of the plastic deformation caused by the autofrettage process and generation of the desirable stress-strain pattern require a much longer period. As shown, a third-order polynomial equation best described the interdependence between the time required for the completion of the process, the corresponding autofrettage pressure and the generated strain state. The method presented can be used as a tool for the determination of the optimal autofrettage process parameters in industrial production of fuel injection lines.

  13. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  14. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

  15. Results of studying of turbulent heat transfer deterioration and their application for development of engineering methods of calculation of heat transfer and pressure drop in supercritical-pressure coolant flow

    International Nuclear Information System (INIS)

    Vladimir A Kurganov; Yuri A Zeigarnik

    2005-01-01

    Full text of publication follows: Using of the supercritical-pressure (SCP) water as a working medium is an apparent way to increase specific capacity and economic efficiency of nuclear power installations. Nevertheless, to provide safe operation of SCP nuclear power units, it is necessary to considerably improve reliability and accuracy of calculations of pressure drop and heat transfer in the SCP working media and coolants flows and the methods of forecasting such a dangerous phenomenon as deterioration of the turbulent heat transfer at a certain level of heat flux density. A value of the latter changes within a very large range depending on the specific conditions of the process under consideration. In the paper, the main results of the experimental study of heat transfer, pressure drop, and velocity and temperature fields in both upward and downward flows of the SCP CO 2 in tubes are considered. This study was conducted at OIVT RAN under conditions of heat input and embraced the regimes of normal and deteriorated heat transfer as well. On the basis of this data, the concept regarding to physical mechanism of incipience of the regimes of deteriorated heat transfer was developed. Classification of different modes of heat transfer deterioration in vertical channels is proposed. A degree of a danger of certain regimes is assessed. It is shown that the above phenomenon is caused by transformation of the structure of nonisothermal flow of SCP fluid due to changes in proportions between the forces acting upon a flow, specifically, because of an increase in the inertia forces due to thermal acceleration of a flow and/or in Archimedes' (buoyancy) forces up to the level comparable or higher than that of friction forces. The efficiency of the most thorough correlations for calculating normal and deteriorated heat transfer in flows of SCP water and CO 2 is analyzed. Reliability of existed recommendations to determine boundaries of normal heat transfer regimes is considered

  16. Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

    International Nuclear Information System (INIS)

    Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

    1995-01-01

    The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

  17. High repetition rate driver circuit for modulation of injection lasers

    International Nuclear Information System (INIS)

    Dornan, B.R.; Goel, J.; Wolkstein, H.J.

    1981-01-01

    An injection laser modulator comprises a self-biased field effect transistor (FET) and an injection laser to provide a quiescent state during which lasing of the injection laser occurs in response to a high repetition rate signal of pulse coded modulation (pcm). The modulator is d.c. coupled to an input pulse source of pcm rendering it compatible with an input pulse referenced to ground and not being subject to voltage level shifting of the input pulse. The modulator circuit in its preferred and alternate embodiments provides various arrangements for high impedance input and low impedance output matching. In addition, means are provided for adjusting the bias of the FET as well as the bias of the injection laser

  18. HIGH BLOOD PRESSURE: DOES THIS CONCERN ME?

    CERN Multimedia

    2007-01-01

    To find out, the Medical Service's nurses are organising A HIGH BLOOD PRESSURE SCREENING AND PREVENTION CAMPAIGN from Monday, 26th to Thursday, 29th March 2007 at the Infirmary - Building 57 - ground floor A blood pressure test, advice, information and, if necessary, referral for specialist medical treatment will be offered to any person working on the CERN site. High blood pressure is a silent threat to health. So come and get your blood pressure checked.

  19. HIGH BLOOD PRESSURE: DOES THIS CONCERN ME?

    CERN Multimedia

    2007-01-01

    To find out, the Medical Service's nurses are organising A HIGH BLOOD PRESSURE SCREENING AND PREVENTION CAMPAIGN from Monday, 26th to Thursday, 29th March 2007 at the Infirmary - Building 57 - ground floor A blood pressure test, advice, information and, if necessary, referral for specialist medical treatment will be offered to any person working on the CERN site. High blood pressure is a stealth threat to health. So come and get your blood pressure checked.

  20. Environmental response nanosilica for reducing the pressure of water injection in ultra-low permeability reservoirs

    Science.gov (United States)

    Liu, Peisong; Niu, Liyong; Li, Xiaohong; Zhang, Zhijun

    2017-12-01

    The super-hydrophobic silica nanoparticles are applied to alter the wettability of rock surface from water-wet to oil-wet. The aim of this is to reduce injection pressure so as to enhance water injection efficiency in low permeability reservoirs. Therefore, a new type of environmentally responsive nanosilica (denote as ERS) is modified with organic compound containing hydrophobic groups and "pinning" groups by covalent bond and then covered with a layer of hydrophilic organic compound by chemical adsorption to achieve excellent water dispersibility. Resultant ERS is homogeneously dispersed in water with a size of about 4-8 nm like a micro-emulsion system and can be easily injected into the macro or nano channels of ultra-low permeability reservoirs. The hydrophobic nanosilica core can be released from the aqueous delivery system owing to its strong dependence on the environmental variation from normal condition to injection wells (such as pH and salinity). Then the exposed silica nanoparticles form a thin layer on the surface of narrow pore throat, leading to the wettability from water-wet to oil-wet. More importantly, the two rock cores with different permeability were surface treated with ERS dispersion with a concentration of 2 g/L, exhibit great reduce of water injection pressure by 57.4 and 39.6%, respectively, which shows great potential for exploitation of crude oil from ultra-low permeability reservoirs during water flooding. [Figure not available: see fulltext.

  1. ''Iodine delivery rate'' with catheterangiography under pressure conditions of hand injection

    International Nuclear Information System (INIS)

    Busch, H.P.; Stocker, K.P.

    1998-01-01

    Purpose: The aim of this study was to record the flow-rate and to calculate the 'iodine delivery rate' (IDR) of contrast media of various viscosities when the contrast media are injected by hand. Methods: Five different catheters for coronary angiography were tested with the injection system Medral Mark V Plus. Injections were performed with pressures of 100, 200 and 400 PSI. The contrast media examined were Imeron 350, Imeron 400, Omnipaque 350 and Ultravist 370. The IDR was calculated on the basis of the measured flow rate and the Iodine content of the contrast medium. Results: As was expected, the IDR was higher as the pressure increased. In addition to the iodine content the viscosity of the contrast medium is a very important factor for the IDR. At both 100 PSI and 200 PSI the increase of the IDR was higher with Imeron 350 than with Imeron 400. The comparison of contrast media with identical iodine contents but differing viscosities (Imeron 350, Omnipaque 350) clearly showed that the IDR depended on viscosity. Conclusion: The 'iodine delivery rate' is a decisive factor in the opacification of arterial vessels. Both iodine content and viscosity of a contrast medium are important for the IDR. Because of the low pressure at manual injection, contrast media with low viscosities should be used. A further possibility to increase the IDR is warming-up the contrast medium to body temperature. (orig.) [de

  2. High pressure effect for high-Tc superconductors

    International Nuclear Information System (INIS)

    Takahashi, Hiroki; Tomita, Takahiro

    2011-01-01

    A number of experimental and theoretical studies have been performed to understand the mechanism of high-T c superconductivity and to enhance T c . High-pressure techniques have played a very important role for these studies. In this paper, the high-pressure techniques and physical properties of high-T c superconductor under high pressure are presented. (author)

  3. Flooding-limited thermal mixing: The case of high-froude number injection

    International Nuclear Information System (INIS)

    Iyer, K.; Theofanous, T.G.

    1985-01-01

    The stratification in the cold leg due to high pressure injection in a stagnated loop of a PWR is considered. The working hypothesis is that at high injection Froude numbers the extent of mixing approaches a limit controlled only by the flooding condition at the cold leg exit. The available experimental data support this hypothesis. Predictions for reactor conditions indicate a stratification of about --40 0 C. As a consequence, the downcomer plume would be rather weak (low Froude Number) and is expected to decay quickly

  4. 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

  5. A control-oriented approach to estimate the injected fuel mass on the basis of the measured in-cylinder pressure in multiple injection diesel engines

    International Nuclear Information System (INIS)

    Finesso, Roberto; Spessa, Ezio

    2015-01-01

    Highlights: • Control-oriented method to estimate injected quantities from in-cylinder pressure. • Able to calculate the injected quantities for multiple injection strategies. • Based on the inversion of a heat-release predictive model. • Low computational time demanding. - Abstract: A new control-oriented methodology has been developed to estimate the injected fuel quantities, in real-time, in multiple injection DI diesel engines on the basis of the measured in-cylinder pressure. The method is based on the inversion of a predictive combustion model that was previously developed by the authors, and that is capable of estimating the heat release rate and the in-cylinder pressure on the basis of the injection rate. The model equations have been rewritten in order to derive the injected mass as an output quantity, starting from use of the measured in-cylinder pressure as input. It has been verified that the proposed method is capable of estimating the injected mass of pilot pulses with an uncertainty of the order of ±0.15 mg/cyc, and the total injected mass with an uncertainty of the order of ±0.9 mg/cyc. The main sources of uncertainty are related to the estimation of the in-cylinder heat transfer and of the isentropic coefficient γ = c_p/c_v. The estimation of the actual injected quantities in the combustion chamber can represent a powerful means to diagnose the behavior of the injectors during engine operation, and offers the possibility of monitoring effects, such as injector ageing and injector coking, as well as of allowing an accurate control of the pilot injected quantities to be obtained; the latter are in fact usually characterized by a large dispersion, with negative consequences on the combustion quality and emission formation. The approach is characterized by a very low computational time, and is therefore suitable for control-oriented applications.

  6. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

    In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

  7. Pressure pressure-balanced pH sensing system for high temperature and high pressure water

    International Nuclear Information System (INIS)

    Tachibana, Koji

    1995-01-01

    As for the pH measurement system for high temperature, high pressure water, there have been the circumstances that first the reference electrodes for monitoring corrosion potential were developed, and subsequently, it was developed for the purpose of maintaining the soundness of metallic materials in high temperature, high pressure water in nuclear power generation. In the process of developing the reference electrodes for high temperature water, it was clarified that the occurrence of stress corrosion cracking in BWRs is closely related to the corrosion potential determined by dissolved oxygen concentration. As the types of pH electrodes, there are metal-hydrogen electrodes, glass electrodes, ZrO 2 diaphragm electrodes and TiO 2 semiconductor electrodes. The principle of pH measurement using ZrO 2 diaphragms is explained. The pH measuring system is composed of YSZ element, pressure-balanced type external reference electrode, pressure balancer and compressed air vessel. The stability and pH response of YSZ elements are reported. (K.I.)

  8. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

  9. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

  10. Continuous surveillance of reactor coolant circuit integrity

    International Nuclear Information System (INIS)

    1986-01-01

    Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

  11. Multidimensional modeling of the effect of fuel injection pressure on temperature distribution in cylinder of a turbocharged DI diesel engine

    Directory of Open Access Journals (Sweden)

    Sajjad Emami

    2013-06-01

    Full Text Available In this study, maintaining a constant fuel rate, injection pressure of 275 bar to 1000 bar (275×102 kPa to 1000×102 kPa, has been changed. Effect of injection pressure, the pressure inside the cylinder on the free energy, power, engine indicators, particularly indicators of fuel consumption, pollutants and their effects on parameters affecting the output of the engine combustion chamber have been studied in droplet diameter. Finally, the effects of fuel mixture equivalence, Cantor temperature, soot and NOx due to the increase of injection pressure, engine efficiency and emissions have been examined.

  12. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  13. Fabrication of High Temperature and High Pressure Vessel for the Fuel Test

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Sim, Bong Shick; Shon, Jae Min; Ahn, Seung Ho; Yoo, Seong Yeon

    2007-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR and CANDU nuclear power plants has been developed and installed in HANARO, KAERI. It is consisted of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS which is located inside the pool is divided into 3-parts; they are in-pool pipes, IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The localization of the IVA is achieved by manufacturing through local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique of the instrument lines has been checked for its functionality and yield. A IVA has been manufactured by local technique and will be finally tested under out of the high temperature and high pressure test

  14. High-pressure boron hydride phases

    International Nuclear Information System (INIS)

    Barbee, T.W. III; McMahan, A.K.; Klepeis, J.E.; van Schilfgaarde, M.

    1997-01-01

    The stability of boron-hydrogen compounds (boranes) under pressure is studied from a theoretical point of view using total-energy methods. We find that the molecular forms of boranes known to be stable at ambient pressure become unstable at high pressure, while structures with extended networks of bonds or metallic bonding are energetically favored at high pressures. If such structures are metastable on return to ambient pressure, they would be energetic as well as dense hydrogen storage media. An AlH 3 -like structure of BH 3 is particularly interesting in that it may be accessible by high-pressure diamond anvil experiments, and should exhibit both second-order structural and metal-insulator transitions at lower pressures. copyright 1997 The American Physical Society

  15. High-rate injection is associated with the increase in U.S. mid-continent seismicity

    Science.gov (United States)

    Weingarten, Matthew; Ge, Shemin; Godt, Jonathan W.; Bekins, Barbara A.; Rubinstein, Justin L.

    2015-01-01

    An unprecedented increase in earthquakes in the U.S. mid-continent began in 2009. Many of these earthquakes have been documented as induced by wastewater injection. We examine the relationship between wastewater injection and U.S. mid-continent seismicity using a newly assembled injection well database for the central and eastern United States. We find that the entire increase in earthquake rate is associated with fluid injection wells. High-rate injection wells (>300,000 barrels per month) are much more likely to be associated with earthquakes than lower-rate wells. At the scale of our study, a well’s cumulative injected volume, monthly wellhead pressure, depth, and proximity to crystalline basement do not strongly correlate with earthquake association. Managing injection rates may be a useful tool to minimize the likelihood of induced earthquakes.

  16. Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

    International Nuclear Information System (INIS)

    Sharma, V.K.; Surendar, C.; Bapat, C.N.

    1994-01-01

    The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

  17. Evaluation of a postulated loss of coolant accident (LOCA) due to a 160 cm2 break in a cold leg of Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Azevedo, Carlos Vicente Goulart de; Palmieri, Elcio Tadeu; Aronne, Ivan Dionysio

    2002-01-01

    The development of a qualified full nodalization of Angra2 NPP for RELAP5/Mod 3.2.2 gamma, aiming at the evaluation of a comprehensive number of accidents and transients, thus providing suitable safety analysis support for licensing purposes, is being carried out within the framework of CNEN internal technical cooperation, involving some of its institutes (CDTN, IPEN and IEN) and the Reactors Coordination (CODRE). This work presents a simulation of a postulated Angra2 small cold leg break loss of coolant accident (SBLOCA). A 160 cm 2 break is supposed to occur at one cold leg between the main coolant pump and the reactor vessel and is described in the Angra2 Final Safety Analysis Report, section 15.6.4.1.3.4. The simulation of several types of transients and accidents is necessary to verify the adequate performance of the modeled logic and systems. In general, the analysis of such and accident allows to demonstrate the safety Injection System performance and the reliable transition between the high pressure safety injection, the accumulator injection and the residual heat removal phases. Furthermore, it is assumed that some components are out of service due to fail or repair in order to make a conservative analysis. The results showed a compatible behavior of the molded systems and that the simulated Emergency Core Cooling System was able to provide sufficient cooling to avoid any damage to the core. (author)

  18. Heating efficiency of high-power perpendicular neutral-beam injection in PDX

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Arunasalam, V.; Bell, M.

    1982-03-01

    The heating efficiency of high power (up to 7.2 MW) near-perpendicular neutral beam injection in the PDX tokamak is comparable to that of tangential injection in PLT. Collisionless plasmas with central ion temperatures up to 6.5 keV and central electron temperatures greater than 2.5 keV have been obtained. The plasma pressure, including the contribution from the beam particles, increases with increasing beam power and does not appear to saturate, although the parametric dependence of the energy confinement time is different from that observed in ohmic discharges

  19. SATCAP-C : a program for thermal hydraulic design of pressurized water injection type capsule

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Someya, Hiroyuki; Asoh, Tomokazu; Niimi, Motoji

    1992-10-01

    There are capsules called 'Pressure Water Injection Type Capsule' as a kind of irradiation devices at the Japan Materials Testing Reactor (JMTR). A type of the capsules is a 'Boiling Water Capsule' (usually named BOCA). The other type is a 'Saturated Temperature Capsule' (named SATCAP). When the water is kept at a constant pressure, the water temperature does not become higher than the saturated temperature so far as the water does not fully change to steam. These type capsules are designed on the basis of the conception of applying the water characteristic to the control of irradiation temperature of specimens in the capsules. In designing of the capsules in which the pressurized water is injected, thermal performances have to be understood as exactly as possible. It is not easy however to predict thermal performances such as axially temperature distribution of water injected in the capsule, because there are heat-sinks at both side of inner and outer of capsule casing as the result that the water is fluid. Then, a program (named SATCAP-C) for the BOCA and SATCAP was compiled to grasp the thermal performances in the capsules and has been used the design of the capsules and analysis of the data obtained from some actual irradiation capsules. It was confirmed that the program was effective in thermal analysis for the capsules. The analysis found out the values for heat transfer coefficients at various surfaces of capsule components and some thermal characteristics of capsules. (author)

  20. A highly efficient six-stroke internal combustion engine cycle with water injection for in-cylinder exhaust heat recovery

    International Nuclear Information System (INIS)

    Conklin, James C.; Szybist, James P.

    2010-01-01

    A concept adding two strokes to the Otto or Diesel engine cycle to increase fuel efficiency is presented here. It can be thought of as a four-stroke Otto or Diesel cycle followed by a two-stroke heat recovery steam cycle. A partial exhaust event coupled with water injection adds an additional power stroke. Waste heat from two sources is effectively converted into usable work: engine coolant and exhaust gas. An ideal thermodynamics model of the exhaust gas compression, water injection and expansion was used to investigate this modification. By changing the exhaust valve closing timing during the exhaust stroke, the optimum amount of exhaust can be recompressed, maximizing the net mean effective pressure of the steam expansion stroke (MEP steam ). The valve closing timing for maximum MEP steam is limited by either 1 bar or the dew point temperature of the expansion gas/moisture mixture when the exhaust valve opens. The range of MEP steam calculated for the geometry of a conventional gasoline engine and is from 0.75 to 2.5 bars. Typical combustion mean effective pressures (MEP combustion ) of naturally aspirated gasoline engines are up to 10 bar, thus this concept has the potential to significantly increase the engine efficiency and fuel economy.