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Sample records for low 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. 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)

  5. 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

  6. 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

  7. 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.

  8. 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)

  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

    2003-01-01

    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

  10. 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

  11. 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

  12. 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.)

  13. 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

  14. 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

  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. 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)

  17. 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

  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. 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.)

  20. 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.

  1. 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)

  2. 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)

  3. 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.)

  4. 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.

  5. 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

  6. 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

  7. 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.)

  8. 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.

  9. 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.)

  10. 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.

  11. 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

  12. Experimental Verification of Integrity of Low-Pressure Injection Piles Structure - Pile Internal Capacity

    Science.gov (United States)

    Pachla, Henryk

    2017-12-01

    The idea of strengthening the foundation using injection piles lies in transferring loads from the foundation to the piles anchorage in existing structure and formed in the soil. Such a system has to be able to transfer loads from the foundation to the pile and from the pile onto the soil. Pile structure often reinforced with steel element has to also be able to transfer such a loading. According to the rules of continuum mechanics, the bearing capacity of such a system and a deformation of its individual elements can be determined by way of an analysis of the contact problem of three interfaces. Each of these surfaces is determined by different couples of materials. Those surfaces create: pile-foundation anchorage, bonding between reinforcement and material from which the pile is formed and pilesoil interface. What is essential is that on the contact surfaces the deformation of materials which adhere to each other can vary and depends on the mechanical properties and geometry of these surfaces. Engineering practice and experimental research point out that the failure in such structures occurs at interfaces. The paper is concentrating on presenting the experiments on interaction between cement grout and various types of steel reinforcement. The tests were conducted on the special low pressure injection piles widely used to strengthen foundations of already existing structures of historical buildings due to the technology of formation and injection pressure.

  13. Experimental Verification of Integrity of Low-Pressure Injection Piles Structure – Pile Internal Capacity

    Directory of Open Access Journals (Sweden)

    Pachla Henryk

    2017-12-01

    Full Text Available The idea of strengthening the foundation using injection piles lies in transferring loads from the foundation to the piles anchorage in existing structure and formed in the soil. Such a system has to be able to transfer loads from the foundation to the pile and from the pile onto the soil. Pile structure often reinforced with steel element has to also be able to transfer such a loading. According to the rules of continuum mechanics, the bearing capacity of such a system and a deformation of its individual elements can be determined by way of an analysis of the contact problem of three interfaces. Each of these surfaces is determined by different couples of materials. Those surfaces create: pile-foundation anchorage, bonding between reinforcement and material from which the pile is formed and pilesoil interface. What is essential is that on the contact surfaces the deformation of materials which adhere to each other can vary and depends on the mechanical properties and geometry of these surfaces. Engineering practice and experimental research point out that the failure in such structures occurs at interfaces. The paper is concentrating on presenting the experiments on interaction between cement grout and various types of steel reinforcement. The tests were conducted on the special low pressure injection piles widely used to strengthen foundations of already existing structures of historical buildings due to the technology of formation and injection pressure.

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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)

  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. Modeling the Rapid Boil-Off of a Cryogenic Liquid When Injected into a Low Pressure Cavity

    Science.gov (United States)

    Lira, Eric

    2016-01-01

    Many launch vehicle cryogenic applications require the modeling of injecting a cryogenic liquid into a low pressure cavity. The difficulty of such analyses lies in accurately predicting the heat transfer coefficient between the cold liquid and a warm wall in a low pressure environment. The heat transfer coefficient and the behavior of the liquid is highly dependent on the mass flow rate into the cavity, the cavity wall temperature and the cavity volume. Testing was performed to correlate the modeling performed using Thermal Desktop and Sinda Fluint Thermal and Fluids Analysis Software. This presentation shall describe a methodology to model the cryogenic process using Sinda Fluint, a description of the cryogenic test set up, a description of the test procedure and how the model was correlated to match the test results.

  4. 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

  5. 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

  6. 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.

  7. 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

  8. Physical characterisation of particles and rheological of a heterogeneous system used in low-pressure injection moulding

    International Nuclear Information System (INIS)

    Zampieron, Joao Vicente

    2002-01-01

    The powder injection moulding process is a recent technology, which offers as advantages a high production of complex geometry metal parts, with low cost, where secondary operations of machinery are unnecessary. The main of this thesis was centered on a coarse powders feedstock injection. The process begins with the composition of the mass, that is the combination of metal powders with organic binders. The following steps succeed injection in moulds, debinding, sintering and, if necessary, cleaning. For the formulation of the feedstock it is indispensable the characterisation of the powders. This is little mentioned in the open literature and brings up controversy among authors. At first, a series of powders characterisations of AISI 316 L stainless steel (below 25 μm) was adopted. The next step was to characterise the rheological behaviour of the feedstock using different rheological apparatus, so as to find the most appropriate equipment to the low-pressure powder injection molding process. The mass has to present a favourable rheological behaviour, which is low viscosity. The results of the physical characterisation were correlated among themselves and with the rheological characterisation. This was undertaken with the purpose of finding agreement among their values. Finally, the possibility of injection of water and gas atomised stainless steel coarse powders feedstock was studied. This presents as main advantage, the reduction of costs for the process. According to the literature, only powders with size below 25 μm are possible to be injected. Hence, starting from the physical characterisation of particles and rheological characterisation of the feedstock, the formulation of an appropriate mass was found for the coarse powders. These coarse powders were characterised by particles below 45 μm. In this case it was necessary to alter drastically the feedstock composition, using high amounts of wax, which lead to unstable rheological conditions. But, it was

  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. Continuous positive airway pressure breathing increases the spread of sensory blockade after low-thoracic epidural injection of lidocaine.

    NARCIS (Netherlands)

    Visser, W.A.; Gielen, M.J.M.; Giele, J.L.P.

    2006-01-01

    Factors affecting the distribution of sensory blockade after epidural injection of local anesthetics remain incompletely clarified. To evaluate if increasing intrathoracic pressure affects the spread of thoracic epidural anesthesia, we randomized 20 patients who received an epidural catheter at the

  11. 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.

  12. 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

  13. 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

  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. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  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. 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. 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. 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

  20. Low Blood Pressure (Hypotension)

    Science.gov (United States)

    ... lowest at night and rises sharply on waking. Blood pressure: How low can you go? What's considered low ... low blood pressure. Medications that can cause low blood pressure Some medications can cause low blood pressure, including: ...

  1. 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

  2. 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

  3. 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. 

  4. 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.

  5. 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

  6. Low Blood Pressure

    Science.gov (United States)

    ... a problem. Sometimes blood pressure that is too low can also cause problems. Blood pressure is the ... reading is 90/60 or lower, you have low blood pressure. Some people have low blood pressure ...

  7. 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

  8. 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

  9. 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

  10. 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.

  11. Low-activation lead coolant for advanced small modular NPP

    International Nuclear Information System (INIS)

    Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

    2001-01-01

    The purpose of the paper is in studying perspectives of a new heavy liquid metal coolant for a small fast reactor (FR) concept. To reduce the post irradiation activity of the coolant the using of lead isotope, Pb-206, instead of natural lead, Pb-nat, is offered. In this case the accumulation of such hazardous radionuclides, as Po-210, Bi-208, Bi-207, essentially decreases. The interval of the lead-206 coolant cost which does not exceed 20% of the overall FR cost is estimated. The possibility of lead-206 obtaining for FR needs with the centrifugal separation technique is pointed out. (author)

  12. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    International Nuclear Information System (INIS)

    Chopra, O.K.; Shack, W.J.

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented

  13. 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.

  14. 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.)

  15. 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

  16. 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.

  17. Comparison of MELCOR modeling techniques and effects of vessel water injection on a low-pressure, short-term, station blackout at the Grand Gulf Nuclear Station

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1995-06-01

    A fully qualified, best-estimate MELCOR deck has been prepared for the Grand Gulf Nuclear Station and has been run using MELCOR 1.8.3 (1.8 PN) for a low-pressure, short-term, station blackout severe accident. The same severe accident sequence has been run with the same MELCOR version for the same plant using the deck prepared during the NUREG-1150 study. A third run was also completed with the best-estimate deck but without the Lower Plenum Debris Bed (BH) Package to model the lower plenum. The results from the three runs have been compared, and substantial differences have been found. The timing of important events is shorter, and the calculated source terms are in most cases larger for the NUREG-1150 deck results. However, some of the source terms calculated by the NUREG-1150 deck are not conservative when compared to the best-estimate deck results. These results identified some deficiencies in the NUREG-1150 model of the Grand Gulf Nuclear Station. Injection recovery sequences have also been simulated by injecting water into the vessel after core relocation started. This marks the first use of the new BH Package of MELCOR to investigate the effects of water addition to a lower plenum debris bed. The calculated results indicate that vessel failure can be prevented by injecting water at a sufficiently early stage. No pressure spikes in the vessel were predicted during the water injection. The MELCOR code has proven to be a useful tool for severe accident management strategies

  18. 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)

  19. 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...

  20. 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)

  1. Attempt of lean burn of a 4 cycle gasoline engine by the aid of low pressure air assisted in-cylinder injection; Tonai kuki nenryo funsha ni yoru lean burn no kokoromi

    Energy Technology Data Exchange (ETDEWEB)

    Hatakeyama, S; Kondo, M; Sekiya, Y; Murayama, T [Hokkaido Automotive Engineering College, Hokkaido (Japan)

    1997-10-01

    Comparable performance and exhaust emission with conventional carburetor was obtained by a low Pressure air assisted in-cylinder injection system. And lean burn of idling and light load operation till A/F=70 was realized by installing a spark Plug and a reed type injection nozzle in a divided combustion chambaer of a 4 cycle gasoline engine. 2 refs., 10 figs.

  2. 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)

  3. 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

  4. 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.)

  5. 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.)

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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)

  14. 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.

  15. 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.)

  16. 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)

  17. 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.

  18. 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.)

  19. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C.

    2012-10-01

    The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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.

  5. Investigation of low pressure ES-SAGD

    Energy Technology Data Exchange (ETDEWEB)

    Ivory, J.; Zheng, R.; Nasr, T.; Deng, X.; Beaulieu, G.; Heck, G. [Alberta Research Council, Edmonton, AB (Canada)

    2008-10-15

    This paper described a scaled model experiment conducted to investigate the effectiveness of expanding solvent steam assisted gravity drainage (ES-SAGD) processes at low pressures. Lower SAGD pressures typically result in reduced oil production as a result of correspondingly lower steam temperatures. However, lower pressures may also result in a reduced steam to oil ratio (SOR) and a higher vaporization heat. Steam was injected into an injection well at 33 cm{sup 3} per minute and in a production well at 31 cm{sup 3} per minute. Steam and solvents were then co-injected into the injection well at a temperature of 206 degrees C. The experiment was history-matched and a parametric analysis was conducted using a simulation tool. The 2-D and 3-D field-scale simulations investigated the impact of operating pressures, injection rates; sub-cool; oil and gas phase diffusion and dispersion; live oil versus dead oil performance; and the use of drawdown when oil rates declined. Low pressure ES-SAGD was then compared with low-pressure SAGD. Results of the study suggested that production pressures, sub-cool and solvent concentrations are important parameters in ES-SAGD processes. At 1500 kPa production pressure and 10 degrees C sub-cool, the co-injection of solvent with steam increased average oil rates by 15 per cent more than the SAGD process. SOR was also reduced. 6 refs., 8 tabs., 20 figs.

  6. 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.)

  7. 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)

  8. 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.

  9. 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)

  10. 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

  11. 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.

  12. Pressure drops in low pressure local boiling

    International Nuclear Information System (INIS)

    Courtaud, Michel; Schleisiek, Karl

    1969-01-01

    For prediction of flow reduction in nuclear research reactors, it was necessary to establish a correlation giving the pressure drop in subcooled boiling for rectangular channels. Measurements of pressure drop on rectangular channel 60 and 90 cm long and with a coolant gap of 1,8 and 3,6 mm were performed in the following range of parameters. -) 3 < pressure at the outlet < 11 bars abs; -) 25 < inlet temperature < 70 deg. C; -) 200 < heat flux < 700 W/cm 2 . It appeared that the usual parameter, relative length in subcooled boiling, was not sufficient to correlate experimental pressure losses on the subcooled boiling length and that there was a supplementary influence of pressure, heat flux and subcooling. With an a dimensional parameter including these terms a correlation was established with an error band of ±10%. With a computer code it was possible to derive the relation giving the overall pressure drop along the channel and to determine the local gradients of pressure drop. These local gradients were then correlated with the above parameter calculated in local conditions. 95 % of the experimental points were computed with an accuracy of ±10% with this correlation of gradients which can be used for non-uniform heated channels. (authors) [fr

  13. 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.)

  14. 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

  15. 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

  16. 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

  17. 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.

  18. Low pressure lithium condensation

    International Nuclear Information System (INIS)

    Wadkins, R.P.; Oh, C.H.

    1985-01-01

    A low pressure experiment to evaluate the laminar film condensation coefficients of lithium was conducted. Some thirty-six different heat transfer tests were made at system pressures ranging from 1.3 to 26 Pa. Boiled lithium was condensed on the inside of a 7.6-cm (ID), 409 stainless-steel pipe. Condensed lithium was allowed to reflux back to the pool boiling region below the condensing section. Fourteen chromel/alumel thermocouples were attached in various regions of the condensing section. The thermocouples were initially calibrated with errors of less than one degree Celsius

  19. 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

  20. 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)

  1. Low-pressure

    Energy Technology Data Exchange (ETDEWEB)

    Baker, Richard [Membrane Technology And Research, Inc., Newark, CA (United States); Kniep, Jay [Membrane Technology And Research, Inc., Newark, CA (United States); Hao, Pingjiao [Membrane Technology And Research, Inc., Newark, CA (United States); Chan, Chi Cheng [Membrane Technology And Research, Inc., Newark, CA (United States); Nguyen, Vincent [Membrane Technology And Research, Inc., Newark, CA (United States); Huang, Ivy [Membrane Technology And Research, Inc., Newark, CA (United States); Amo, Karl [Membrane Technology And Research, Inc., Newark, CA (United States); Freeman, Brice [Membrane Technology And Research, Inc., Newark, CA (United States); Fulton, Don [Membrane Technology And Research, Inc., Newark, CA (United States); Ly, Jennifer [Membrane Technology And Research, Inc., Newark, CA (United States); Lipscomb, Glenn [Membrane Technology And Research, Inc., Newark, CA (United States); Lou, Yuecun [Membrane Technology And Research, Inc., Newark, CA (United States); Gogar, Ravikumar [Membrane Technology And Research, Inc., Newark, CA (United States)

    2015-01-29

    This final technical progress report describes work conducted by Membrane Technology and Research, Inc. (MTR) for the Department of Energy (DOE NETL) on development of low-pressure membrane contactors for carbon dioxide (CO2) capture from power plant flue gas (award number DE-FE0007553). The work was conducted from October 1, 2011 through September 30, 2014. The overall goal of this three-year project was to build and operate a prototype 500 m2 low-pressure sweep membrane module specifically designed to separate CO2 from coal-fired power plant flue gas. MTR was assisted in this project by a research group at the University of Toledo, which contributed to the computational fluid dynamics (CFD) analysis of module design and process simulation. This report details the work conducted to develop a new type of membrane contactor specifically designed for the high-gas-flow, low-pressure, countercurrent sweep operation required for affordable membrane-based CO2 capture at coal power plants. Work for this project included module development and testing, design and assembly of a large membrane module test unit at MTR, CFD comparative analysis of cross-flow, countercurrent, and novel partial-countercurrent sweep membrane module designs, CFD analysis of membrane spacers, design and fabrication of a 500 m2 membrane module skid for field tests, a detailed performance and cost analysis of the MTR CO2 capture process with low-pressure sweep modules, and a process design analysis of a membrane-hybrid separation process for CO2 removal from coal-fired flue gas. Key results for each major task are discussed in the report.

  2. 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.)

  3. 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

  4. 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.

  5. 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)

  6. DC injection into low voltage AC networks

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This report summarises the results of a study investigating the impact of levels of injected DC current injections on a low voltage AC distribution network systems in order to recommend acceptable limits of DC from microgeneration. Relevant literature is reviewed, and the impact of DC levels in distribution transformers, transformer modelling, and instrumental transformers are discussed. The impact of DC in residual current devices (RCD) and in domestic electricity watt hour meters is examined along with DC enhanced corrosion, corrosion failure, and the measurement of DC current injection. Sources of DC injection outlined include DC from computer power supplies, network faults, geomagnetic phenomena, lighting circuits/dimmers, and embedded generators.

  7. 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)

  8. 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 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 ...

  9. 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.

  10. 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.

  11. Physical characterisation of particles and rheological of a heterogeneous system used in low-pressure injection moulding; Caracterizacao fisica de particulas e reologica de um sistema heterogeneo utilizado em moldagem de pos por injecao a baixa pressao

    Energy Technology Data Exchange (ETDEWEB)

    Zampieron, Joao Vicente

    2002-07-01

    The powder injection moulding process is a recent technology, which offers as advantages a high production of complex geometry metal parts, with low cost, where secondary operations of machinery are unnecessary. The main of this thesis was centered on a coarse powders feedstock injection. The process begins with the composition of the mass, that is the combination of metal powders with organic binders. The following steps succeed injection in moulds, debinding, sintering and, if necessary, cleaning. For the formulation of the feedstock it is indispensable the characterisation of the powders. This is little mentioned in the open literature and brings up controversy among authors. At first, a series of powders characterisations of AISI 316 L stainless steel (below 25 {mu}m) was adopted. The next step was to characterise the rheological behaviour of the feedstock using different rheological apparatus, so as to find the most appropriate equipment to the low-pressure powder injection molding process. The mass has to present a favourable rheological behaviour, which is low viscosity. The results of the physical characterisation were correlated among themselves and with the rheological characterisation. This was undertaken with the purpose of finding agreement among their values. Finally, the possibility of injection of water and gas atomised stainless steel coarse powders feedstock was studied. This presents as main advantage, the reduction of costs for the process. According to the literature, only powders with size below 25 {mu}m are possible to be injected. Hence, starting from the physical characterisation of particles and rheological characterisation of the feedstock, the formulation of an appropriate mass was found for the coarse powders. These coarse powders were characterised by particles below 45 {mu}m. In this case it was necessary to alter drastically the feedstock composition, using high amounts of wax, which lead to unstable rheological conditions. But, it

  12. 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)

  13. 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

  14. 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.)

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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.

  1. 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.

  2. 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.

  3. 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)

  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. 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 ...

  6. 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.

  7. 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.

  8. 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)

  9. 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)

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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.)

  15. 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...

  16. 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)

  17. 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.

  18. 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.

  19. 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.)

  20. 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)

  1. 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

  2. Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (I-IV), Part II, IZ-240-0379-1963, Vol. II Head of the low temperature RA reactor coolant loop

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. maximum generated heat in the heads of the coolant loop is 10500 kcal/h and minimum generated heat is 1500 kcal/h. The loops are cooled by CO 2 gas, coolant flow is 420 kg/h, and the pressure is 4.5 atu. There is a need to design and construct the secondary coolant loop for the low temperature coolant loop. This volume includes technical specifications of the secondary CO 2 loop with instructions for construction and testing; needed calculations; specification of materials; cost estimation for materials, equipment and construction; and graphical documentation [sr

  3. 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)

  4. 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

  5. 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)

  6. 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.

  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. 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

  9. 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.

  10. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

    Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

  11. 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.

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. 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.

  19. 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

  20. Development of a feed-and-bleed operation strategy with hybrid-SIT under low pressure condition of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, In Seop, E-mail: jeoni@rpi.edu [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States); Han, Sang Hoon, E-mail: shhan2@kaeri.re.kr [Advanced Research Group, Korea Atomic Energy Research Institute, 70 Daedeok-daero 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Kang, Sang Hee, E-mail: sanghee.kang@khnp.co.kr [NSSS Design Group, Korea Hydro & Nuclear Power Co., Ltd., Central Research Institute, 70, 1312-beongil, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States)

    2017-04-01

    Highlights: • The novel F&B operation strategy with H-SIT and LPSI is developed. • The effectiveness of the H-SITs is verified using thermo-hydraulic simulations. • Success criteria considered for the new F&B operation strategy is identified. • A PSA model of APR+ reflecting the new F&B strategy with H-SIT is developed. • A risk analysis of the proposed F&B operation strategy is performed. - Abstract: While safety functions in current nuclear power plants are mainly provided by active safety systems, recently passive safety systems are being combined with the active systems to strengthen accident mitigation capability and therefore enhance overall plant safety. To this end, securing long-term cooling of the core is of particular importance. This study considers the hybrid safety injection tank (H-SIT), a passive injection system, as a target component to develop a long-term cooling strategy using active and passive systems concurrently. In the feed-and-bleed (F&B) operation, one of the important long-term cooling strategies to maintain core safety in pressurized water reactors, low pressure safety injection (LPSI) pumps are typically considered inoperable as depressurization is first required, which leads to core dry-out before reaching LPSI operable pressure. This study investigates whether H-SITs, with the important design feature of passive coolant injection under any pressure condition of the primary coolant system, can make up the core during depressurization thereby allowing LPSI pumps to be used in F&B operation as an additional means of long-term cooling. The effectiveness of the H-SITs is verified using thermal-hydraulic simulations, and based on the results a novel F&B operation strategy with H-SITs and LPSI pumps is developed. A probabilistic safety assessment (PSA) model is then developed in order to assess the risk effect of the suggested strategy. PSA results demonstrate that the proposed strategy lowers core damage frequency in the target

  1. 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

  2. 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.

  3. 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.

  4. 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)

  5. 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))

  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. 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

  8. 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.)

  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 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.)

  11. 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

  12. 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.

  13. Sequential Injection Chromatography with an Ultra-short Monolithic Column for the Low-Pressure Separation of α-Tocopherol and γ-Oryzanol in Vegetable Oils and Nutrition Supplements.

    Science.gov (United States)

    Thaithet, Sujitra; Kradtap Hartwell, Supaporn; Lapanantnoppakhun, Somchai

    2017-01-01

    A low-pressure separation procedure of α-tocopherol and γ-oryzanol was developed based on a sequential injection chromatography (SIC) system coupled with an ultra-short (5 mm) C-18 monolithic column, as a lower cost and more compact alternative to the HPLC system. A green sample preparation, dilution with a small amount of hexane followed by liquid-liquid extraction with 80% ethanol, was proposed. Very good separation resolution (R s = 3.26), a satisfactory separation time (10 min) and a total run time including column equilibration (16 min) were achieved. The linear working range was found to be 0.4 - 40 μg with R 2 being more than 0.99. The detection limits of both analytes were 0.28 μg with the repeatability within 5% RSD (n = 7). Quantitative analyses of the two analytes in vegetable oil and nutrition supplement samples, using the proposed SIC method, agree well with the results from HPLC.

  14. 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.

  15. 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

  16. Ammonia Synthesis at Low Pressure.

    Science.gov (United States)

    Cussler, Edward; McCormick, Alon; Reese, Michael; Malmali, Mahdi

    2017-08-23

    Ammonia can be synthesized at low pressure by the use of an ammonia selective absorbent. The process can be driven with wind energy, available locally in areas requiring ammonia for synthetic fertilizer. Such wind energy is often called "stranded," because it is only available far from population centers where it can be directly used. In the proposed low pressure process, nitrogen is made from air using pressure swing absorption, and hydrogen is produced by electrolysis of water. While these gases can react at approximately 400 °C in the presence of a promoted conventional catalyst, the conversion is often limited by the reverse reaction, which makes this reaction only feasible at high pressures. This limitation can be removed by absorption on an ammine-like calcium or magnesium chloride. Such alkaline metal halides can effectively remove ammonia, thus suppressing the equilibrium constraints of the reaction. In the proposed absorption-enhanced ammonia synthesis process, the rate of reaction may then be controlled not by the chemical kinetics nor the absorption rates, but by the rate of the recycle of unreacted gases. The results compare favorably with ammonia made from a conventional small scale Haber-Bosch process.

  17. 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

  18. 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)

  19. 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

  20. 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)

  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. 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

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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.)

  9. 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

  10. 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.)

  11. 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

  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. 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

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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.

  20. 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

  1. 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.)

  2. 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

  3. 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.

  4. 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.)

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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.

  10. 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

  11. Modulated diesel fuel injection strategy for efficient-clean utilization of low-grade biogas

    International Nuclear Information System (INIS)

    Wang, Xiaole; Qian, Yong; Zhou, Qiyan; Lu, Xingcai

    2016-01-01

    Highlights: • Influences of direct injection strategy on biogas RCCI mode are researched. • Excessive early pilot injection timing leads to the retard of combustion. • Overall indicated thermal efficiency of low-grade biogas can be higher than 40%. • Pilot injection timing has strong influence on particle size distribution. • Composition of biogas has a great influence on the gas emissions. - Abstract: Recently, as a kind of renewable fuel, low-grade biogas has been researched to apply in internal combustion engine. In this paper, an experimental study was conducted to study the influence of injection strategies on the efficient utilization of low-grade biogas in Reactivity Controlled Compression Ignition (RCCI) mode with port fuel injection of biogas and in-cylinder direct injection of diesel based on a modified electronic controlled high-pressure directly injected compression ignition engine. Considered the high proportion of inert gas in biogas, a four-components simulated gas (H_2:CO:CH_4:N_2 = 5:40:5:50 vol%) has been selected as test fuels to simulate biogas. The effects of several injection control parameters such as pilot injection timing, main injection timing, common rail pressure and pilot injection ratio on the combustion and emissions are analyzed in detail. The research demonstrates that the main injection timing can effectively control the combustion phase and excessive early pilot injection timing leads to retard of combustion. CO emissions are relatively high due to homogenous charge of biogas. NOx and smoke emissions can be effectively controlled. In RCCI mode, the indicated thermal efficiency of biogas/diesel can reach 40%.

  12. 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

  13. 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

  14. 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

  15. 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)

  16. ''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

  17. 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.

  18. 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

  19. 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

  20. [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.

  1. 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.

  2. 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

  3. Low energy, high power injection in JT-60 NBI

    International Nuclear Information System (INIS)

    Mizuno, Makoto; Dairaku, Masayuki; Horiike, Hiroshi

    1988-05-01

    JT-60 neutral beam injector (JT-60 NBI) is designed to inject 20 MW neutral hydrogen beam at energies of 70 ∼ 100 keV and the injection power decreases significantly at low energies (∼40 keV). For the extention of operation region aiming at the low density plasma heating and achieving H-mode by plasma periphery heating, increment of the injection power at low beam energies was required. The single-stage acceleration system was investigated in advance at the Prototype Injector Unit. From this result, the total injection power of 17 MW at 40 keV, 48 A per source was expected at the JT-60 NBI. This system was adopted in the JT-60 NBI from June, 1987 to July, 1987 and 17.6 MW neutral beam injection power was achieved. In the NB heating experiment, the H-mode transition phenomena was observed in JT-60 plasma. (author)

  4. 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

  5. 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,

  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. 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

  8. 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

  9. LHI (low head safety injection) emergency cooling pump test for the EPR trademark in operation with solid matter loaded water

    International Nuclear Information System (INIS)

    Ganzmann, I.; Schulte, C.

    2010-01-01

    Emergency cooling pumps are essential and indispensable components of the NPP safety philosophy. In case of a loss-of coolant accident solid matter (debris: fibrous insulation material, concrete dust, pigment particles) might be released into the coolant, LHSI (low head safety injection) pumps have to ensure their performance capacity for a certain amount of debris without damage or loss of power. The authors describe the development of a test facility. The LHSI was tested in continuous operation over a time period of 14 days with a debris content of 1500 ppm (90% mineral wool fibers, 3% concrete dust, 3% pigment particles, 4% microporous insulation material). The pump did not show any damage or loss of hydraulic power. Further tests including thermoshock conditions (temperature changes of 160 C) are planned.

  10. 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

  11. 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.)

  12. 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).

  13. 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.)

  14. 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

  15. 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.

  16. 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)

  17. Simulation of low pressure water hammer

    Science.gov (United States)

    Himr, D.; Habán, V.

    2010-08-01

    Numerical solution of water hammer is presented in this paper. The contribution is focused on water hammer in the area of low pressure, which is completely different than high pressure case. Little volume of air and influence of the pipe are assumed in water, which cause sound speed change due to pressure alterations. Computation is compared with experimental measurement.

  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. 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

  1. 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.

  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. 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.

  4. Injection therapy for subacute and chronic low-back pain.

    Science.gov (United States)

    Staal, J Bart; de Bie, Rob; de Vet, Henrica Cw; Hildebrandt, Jan; Nelemans, Patty

    2008-07-16

    The effectiveness of injection therapy for low-back pain is still debatable. Heterogeneity of target tissue, pharmacological agent and dosage generally found in randomized controlled trials (RCTs) points to the need for clinically valid comparisons in a literature synthesis. To determine if injection therapy is more effective than placebo or other treatments for patients with subacute or chronic low-back pain. We updated the search of the earlier systematic review and searched the Cochrane Central Register of Controlled Trials, MEDLINE and EMBASE databases from January 1999 to March 2007 for relevant trials reported in English, French, German, Dutch and Nordic languages. We also screened references from trials identified. RCTs on the effects of injection therapy involving epidural, facet or local sites for subacute or chronic low-back pain were included. Studies which compared the effects of intradiscal injections, prolotherapy or Ozone therapy with other treatments, were excluded unless injection therapy with another pharmaceutical agent (no placebo treatment) was part of one of the treatment arms. Studies about injections in sacroiliac joints and studies evaluating the effects of epidural steroids for radicular pain were also excluded. Two review authors independently assessed the quality of the trials. If study data were clinically and statistically too heterogeneous to perform a meta-analysis, we used a best evidence synthesis to summarize the results. The evidence was classified into five levels (strong, moderate, limited, conflicting or no evidence), taking into account the methodological quality of the studies. 18 trials (1179 participants) were included in this updated review. The injection sites varied from epidural sites and facet joints (i.e. intra-articular injections, peri-articular injections and nerve blocks) to local sites (i.e. tender- and trigger points). The drugs that were studied consisted of corticosteroids, local anesthetics and a variety of

  5. Stress corrosion cracking studies on ferritic low alloy pressure vessel steel - water chemistry and modelling aspects

    International Nuclear Information System (INIS)

    Tipping, P.; Ineichen, U.; Cripps, R.

    1994-01-01

    The susceptibility of low alloy ferritic pressure vessel steels (A533-B type) to stress corrosion cracking (SCC) degradation has been examined using various BWR type coolant chemistries. Fatigue pre-cracked wedge-loaded double cantilever beams and also constantly loaded 25 mm thick compact tension specimens have shown classical SCC attack. The influence of parameters such as dissolved oxygen content, water impurity level and conductivity, material chemical composition (sulphur content) and stress intensity level are discussed. The relevance of SCC as a life-limiting degradation mechanism for low alloy ferritic nuclear power plant PV steel is examined. Some parameters, thought to be relevant for modelling SCC processes in low alloy steels in simulated BWR-type coolant, are discussed. 8 refs., 1 fig., 4 tabs

  6. Effects of pilot injection parameters on low temperature combustion diesel engines equipped with solenoid injectors featuring conventional and rate-shaped main injection

    International Nuclear Information System (INIS)

    D’Ambrosio, S.; Ferrari, A.

    2016-01-01

    Highlights: • The influence of the principal pilot injection parameters is discussed for low-temperature combustion systems. • Swirl-sweep and dwell-time sweep results are combined to analyze soot emissions. • The pilot injection effects are investigated in injection profiles featuring rate-shaped main injections. - Abstract: The potential of pilot injection has been assessed on a low-temperature combustion diesel engine for automotive applications, which was characterized by a reduced compression-ratio, high EGR rates and postponed main injection timings. Dwell time sweeps have been carried out for pilot injections with distinct energizing times under different representative steady-state working conditions of the medium load and speed area of the New European Driving Cycle. The results of in-cylinder analyses of the pressure, heat-release rate, temperature and emissions are presented. Combustion noise has been shown to decrease significantly when the pilot injected mass increases, while it is scarcely affected by the dwell time between the pilot and main injections. The HC, CO and fuel consumption trends, with respect to both the pilot injection dwell time and mass, are in line with those of conventional combustion systems, and in particular decreasing trends occur as the pilot injection energizing time is increased. Furthermore, a reduced sensitivity of NO_x emissions to both dwell time and pilot injected mass has been found, compared to conventional combustion systems. Finally, it has been observed that soot emissions diminish as the energizing time is shortened, and their dependence on dwell time is influenced to a great extent by the presence of local zones with reduced air-to-fuel ratios within the cylinder. A combined analysis of the results of swirl sweeps and dwell time sweeps is here proposed as a methodology for the detection of any possible interference between pilot combustion burned gases and the main injected fuel. The effect of pilot

  7. Experimental study on the performance of the vapor injection refrigeration system with an economizer for intermediate pressures

    Science.gov (United States)

    Moon, Chang-Uk; Choi, Kwang-Hwan; Yoon, Jung-In; Kim, Young-Bok; Son, Chang-Hyo; Ha, Soo-Jung; Jeon, Min-Ju; An, Sang-Young; Lee, Joon-Hyuk

    2018-04-01

    In this study, to investigate the performance characteristics of vapor injection refrigeration system with an economizer at an intermediate pressure, the vapor injection refrigeration system was analyzed under various experiment conditions. As a result, the optimum design data of the vapor injection refrigeration system with an economizer were obtained. The findings from this study can be summarized as follows. The mass flow rate through the compressor increases with intermediate pressure. The compression power input showed an increasing trend under all the test conditions. The evaporation capacity increased and then decreased at the intermediate pressure, and as such, it became maximum at the given intermediate pressure. The increased mass flow rate of the by-passed refrigerant enhanced the evaporation capacity at the low medium pressure range, but the increased saturation temperature limited the subcooling degree of the liquid refrigerant after the application of the economizer when the intermediate pressure kept rising, and degenerated the evaporation capacity. The coefficient of performance (COP) increased and then decreased with respect to the intermediate pressures under all the experiment conditions. Nevertheless, there was an optimum intermediate pressure for the maximum COP under each experiment condition. Therefore, the optimum intermediate pressure in this study was found at -99.08 kPa, which is the theoretical standard medium pressure under all the test conditions.

  8. 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

  9. Device for preventing coolant outflow in a reactor

    International Nuclear Information System (INIS)

    Nemoto, Kiyomitsu; Mochizuki, Keiichi.

    1975-01-01

    Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

  10. A fracture mechanics approach for estimating fatigue crack initiation in carbon and low-alloy steels in LWR coolant environments

    International Nuclear Information System (INIS)

    Park, H. B.; Chopra, O. K.

    2000-01-01

    A fracture mechanics approach for elastic-plastic materials has been used to evaluate the effects of light water reactor (LWR) coolant environments on the fatigue lives of carbon and low-alloy steels. The fatigue life of such steel, defined as the number of cycles required to form an engineering-size crack, i.e., 3-mm deep, is considered to be composed of the growth of (a) microstructurally small cracks and (b) mechanically small cracks. The growth of the latter was characterized in terms of ΔJ and crack growth rate (da/dN) data in air and LWR environments; in water, the growth rates from long crack tests had to be decreased to match the rates from fatigue S-N data. The growth of microstructurally small cracks was expressed by a modified Hobson relationship in air and by a slip dissolution/oxidation model in water. The crack length for transition from a microstructurally small crack to a mechanically small crack was based on studies on small crack growth. The estimated fatigue S-N curves show good agreement with the experimental data for these steels in air and water environments. At low strain amplitudes, the predicted lives in water can be significantly lower than the experimental values

  11. 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

  12. Botulinum toxin injections for low-back pain and sciatica.

    Science.gov (United States)

    Waseem, Zeeshan; Boulias, Chris; Gordon, Allan; Ismail, Farooq; Sheean, Geoffrey; Furlan, Andrea D

    2011-01-19

    Adequate relief from low-back pain (LBP) is not always possible. Emerging evidence suggests a role for botulinum neurotoxin (BoNT) injections in treating pain disorders. Proponents of BoNT suggest its properties can decrease muscle spasms, ischemia and inflammatory markers, thereby reducing pain. To determine the effects of botulinum toxin injections in adults with LBP. We searched CENTRAL (The Cochrane Library 2009, issue 3) and MEDLINE, EMBASE, and CINAHL to August 2009; screened references from included studies; consulted with content experts and Allergan. We included published and unpublished randomised controlled trials without language restrictions We included randomised trials that evaluated BoNT serotypes versus other treatments in patients with non-specific LBP of any duration. Two review authors selected the studies, assessed the risk of bias using the Cochrane Back Review Group criteria, and extracted the data using standardized forms. We performed a qualitative analysis due to lack of data. We excluded evidence from nineteen studies due to non-randomisation, incomplete or unpublished data. We included three randomised trials (N =123 patients). Only one study included patients with chronic non-specific LBP; the other two examined unique subpopulations. Only one of the three trials had a low risk of bias and demonstrated that BoNT injections reduced pain at three and eight weeks and improved function at eight weeks better than saline injections. The second trial showed that BoNT injections were better than injections of corticosteroid plus lidocaine or placebo in patients with sciatica attributed to piriformis syndrome. The third trial concluded that BoNT injections were better than traditional acupuncture in patients with third lumbar transverse process syndrome. Both studies with high risk of bias had several key limitations. Heterogeneity of the studies prevented meta-analysis. There is low quality evidence that BoNT injections improved pain, function

  13. Experimental simulation of low rate primary coolant leaks. For the case of vessel head penetrations affected by through wall cracking

    International Nuclear Information System (INIS)

    You, D.; Feron, D.; Turluer, G.

    2002-01-01

    An experimental simulation of primary coolant leaks was carried out to determine how the composition of the leaking liquid would change. The experiment used the EVA experimental setup, specially designed for quantitatively investigating concentration phenomena driven by evaporation. The test showed that the final composition, obtained from a solution representative of the primary coolant at the beginning of the cycle, is highly concentrated and slightly acid. The experimental results are compared with those obtained using the MULTEQ software. (authors)

  14. Investigating skin penetration depth and shape following needle-free injection at different pressures: A cadaveric study.

    Science.gov (United States)

    Seok, Joon; Oh, Chang Taek; Kwon, Hyun Jung; Kwon, Tae Rin; Choi, Eun Ja; Choi, Sun Young; Mun, Seog Kyun; Han, Seung-Ho; Kim, Beom Joon; Kim, Myeung Nam

    2016-08-01

    amount of pressure used. When injecting 5% MB at 6 bars, the depth of infiltration reached 2.323 mm, while that at 8.5 bars reached 8.906 mm. The area of the hole created by the 5% MB injection was 0.797 mm(2) at 6 bars and 0.242 mm(2) at 8.5 bars. Latex injections reached a depth of 3.480 mm at 6 bars and 7.558 mm at 8.5 bars, and the areas were measured at 1.043 mm(2) (6 bars) and 0.355 mm(2) (8.5 bars). Histological examination showed that the injection penetrated as deep as the superficial musculoaponeurotic system at 6 bars and the masseter muscle at 8.5 bars. When injecting material into the skin using a pneumatic needle-free injector, higher-pressure injections result in a hole with smaller area than lower-pressure injections. The depth and shape of skin penetration vary according to the amount of pressure applied. For materials of low density and viscosity, there is a greater difference in penetration depth according to the degree of pressure. Lasers Surg. Med. 48:624-628, 2016. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

  15. 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

  16. 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

  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. 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)

  2. 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.

  3. 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

  4. 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)

  5. 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

  6. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  7. The effect of coolant quantity on local fuel–coolant interactions in a molten pool

    International Nuclear Information System (INIS)

    Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

    2015-01-01

    Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

  8. 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

  9. 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

  10. 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.

  11. 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

  12. Low cost sonoluminescence experiment in pressurized water

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, L; Insabella, M [LADOP, University of Mar del Plata (Argentina); Bilbao, L [INFIP, University of Buenos Aires and CONICET (Argentina)

    2012-06-19

    We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  13. Low cost sonoluminescence experiment in pressurized water

    International Nuclear Information System (INIS)

    Bernal, L; Insabella, M; Bilbao, L

    2012-01-01

    We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  15. 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.

  16. 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

  17. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  18. 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.)

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. RETRAN analysis of inter-system LOCA within the primary coolant pump

    International Nuclear Information System (INIS)

    Gangadharan, A.; Pratt, G.F.

    1992-01-01

    One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

  5. Negative Ions in low pressure discharges

    NARCIS (Netherlands)

    Stoffels - Adamowicz, E.; Stoffels, W.W.; Vender, D.; Haverlag, M.; Kroesen, G.M.W.; Hoog, de F.J.

    1995-01-01

    Several aspects of negative ions in low pressure discharges are treated. The elementary processes, in which negative ions are produced and destroyed, are summarized. The influence of negative ions on plasma operation is analyzed in terms of transport equations. It is shown that diffusion, electric

  6. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. Study of the Effect of Clay Particles on Low Salinity Water Injection in Sandstone Reservoirs

    Directory of Open Access Journals (Sweden)

    Sina Rezaei Gomari

    2017-03-01

    Full Text Available The need for optimal recovery of crude oil from sandstone and carbonate reservoirs around the world has never been greater for the petroleum industry. Water-flooding has been applied to the supplement primary depletion process or as a separate secondary recovery method. Low salinity water injection is a relatively new method that involves injecting low salinity brines at high pressure similar to conventional water-flooding techniques, in order to recover crude oil. The effectiveness of low salinity water injection in sandstone reservoirs depends on a number of parameters such as reservoir temperature, pressure, type of clay particle and salinity of injected brine. Clay particles present on reservoir rock surfaces adsorb polar components of oil and modify wettability of sandstone rocks to the oil-wet state, which is accountable for the reduced recovery rates by conventional water-flooding. The extent of wettability alteration caused by three low salinity brines on oil-wet sandstone samples containing varying clay content (15% or 30% and type of clay (kaolinite/montmorillonite were analyzed in the laboratory experiment. Contact angles of mica powder and clay mixture (kaolinite/montmorillonite modified with crude oil were measured before and after injection with three low salinity sodium chloride brines. The effect of temperature was also analyzed for each sample. The results of the experiment indicate that samples with kaolinite clay tend to produce higher contact angles than samples with montmorillonite clay when modified with crude oil. The highest degree or extent of wettability alteration from oil-wet to intermediate-wet state upon injection with low salinity brines was observed for samples injected with brine having salinity concentration of 2000 ppm. The increase in temperature tends to produce contact angles values lying in the higher end of the intermediate-wet range (75°–115° for samples treated at 50 °C, while their corresponding

  13. MODTURCCLAS analysis of moderator poison/coolant mixing in the calandria due to a pressure tube/calandria tube guillotine rupture during an overpoisoned guaranteed shutdown state

    International Nuclear Information System (INIS)

    Mackinnon, J.C.; Szymanski, J.K.; Balog, G.

    1996-01-01

    This paper reports the results of a study to investigate moderator poison/coolant mixing due to a guillotine rupture of a fuel channel when the reactor is in an overpoisoned guaranteed shutdown state. The analysis, performed using MODTURC C LAS, allowed for study of the mixing characteristics and the spatial and temporal evolution of the concentration fields. Results for simulated breaks at three channel locations show that the poison in the vessel is quite well mixed throughout the transient, resulting in no extensive regions of low poison concentration. MODTURC C LAS calculations show that at all three break locations investigated, the displacement of poison from the vessel through the relief ducts is less than that calculated by both the simple uniform mixing model and piston mixing model. This result is expected to hold for all break locations in the core. (author)

  14. 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)

  15. 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...

  16. 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

  17. 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%.

  18. Investigations into low pressure methanol synthesis

    DEFF Research Database (Denmark)

    Sharafutdinov, Irek

    The central topic of this work has been synthesis, characterization and optimization of novel Ni-Ga based catalysts for hydrogenation of CO2 to methanol. The overall goal was to search for materials that could be used as a low temperature (and low pressure) methanol synthesis catalyst....... This is required for small scale delocalized methanol production sites, where installation of energy demanding compression units should be avoided. The work was triggered by DFT calculations, which showed that certain bimetallic systems are active towards methanol synthesis from CO2 and H2 at ambient pressure...... containing 5:3 molar ratio of Ni:Ga, the intrinsic activity (methanol production rate per active surface area) is comparable to that of highly optimised Cu/ZnO/Al2O3. Formation of the catalyst was investigated with the aid of in-situ XRD and in-situ XAS techniques. The mechanism of alloying was proposed...

  19. 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)

  20. 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)

  1. Low cost sonoluminescence experiment in pressurized water

    Science.gov (United States)

    Bernal, L.; Insabella, M.; Bilbao, L.

    2012-06-01

    We present a low cost design for demostration and mesurements of light emmision from a sonoluminescence experiment. Using presurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  2. Low Pressure Nuclear Thermal Rocket (LPNTR) concept

    International Nuclear Information System (INIS)

    Ramsthaler, J.H.

    1991-01-01

    A background and a description of the low pressure nuclear thermal system are presented. Performance, mission analysis, development, critical issues, and some conclusions are discussed. The following subject areas are covered: LPNTR's inherent advantages in critical NTR requirement; reactor trade studies; reference LPNTR; internal configuration and flow of preliminary LPNTR; particle bed fuel assembly; preliminary LPNTR neutronic study results; multiple LPNTR engine concept; tank and engine configuration for mission analysis; LPNTR reliability potential; LPNTR development program; and LPNTR program costs

  3. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

    Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

  4. 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

  5. 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

  6. 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

  7. 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.)

  8. Potential for low fracture toughness and lamellar tearing on PWR steam generator and reactor coolant pump supports. Resolution of generic technical activity A-12

    International Nuclear Information System (INIS)

    Snaider, R.P.; Hodge, J.M.; Levin, H.A.; Zudans, J.J.

    1979-10-01

    This report summarizes work performed by the Nuclear Regulatory Commission staff and its contractor, Sandia Laboratories, in the resolution of Generic Technical Activity A-12, ''Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports.'' The report describes the technical issues, the technical studies performed by Sandia describes the technical issues, the technical studies performed by Sandia Laboratories, the NRC staff's technical positions based on these studies, and the staff's plan for implementing its technical positions. It also provides recommendations for further work. The complete technical input from Sandia Laboratories is appended to the report

  9. 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.

  10. 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

  11. 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.)

  12. 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

  13. 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)

  14. Fundamentals for the development of a low-activation lead coolant with isotopic enrichment for advanced nuclear power facilities

    International Nuclear Information System (INIS)

    Khorasanov, G.L.; Blokhin, A.I.

    2002-01-01

    The purpose of this paper is to study the prospects of new coolants for fast reactors and accelerator driven systems. The main focus is on their improvement using the isotopic tailoring technique to reduce post-irradiation activity. Calculations using the FISPACT-3 code show that irradiating natural lead (Pb-nat) for 30 years leads to the accumulation of long-lived toxic radionuclides, 207 Bi, 208 Bi and 210 Pb, which extends the cooling down period to the clearance level. This time can be shortened by using the lead isotope 206 Pb instead of Pb-nat. This substantially decreases the concentration of the most toxic polonium isotope, 210 Po. Calculations for lead activation in the hard proton-neutron ADS spectrum were performed using the CASCADE/SNT code. The time-dependent activity of the 207 Bi produced in Pb-nat and 206 Pb after irradiation for one year with a proton beam having an energy of 0.8 GeV and a current of 30 mA is given. The activity of 207 Bi is decreased by four orders of magnitude when 206 Pb is used instead of natural lead as a coolant for ADS targets. The production of such radiotoxic nuclides as 210 Po is also substantially diminished. (author)

  15. 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...

  16. 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)

  17. Experimental study on low pressure flow instability

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin; Wu Shaorong; Bo Jinhai; Zhang Youjie

    1997-05-01

    The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The flow behavior for a wide range of inlet subcooling, in which the flow undergoes from single phase to two phase, is described in a natural circulation system at low pressure (p = 0.1, 0.24 MPa). Several kinds of flow instability, e.g. subcooled boiling instability, subcooled boiling induced flashing instability, pure flashing instability as well as flashing coupled density wave instability and high frequency flow oscillation, are investigated. The mechanism of flashing and flashing concerned flow instability, which has never been studied well in this field, is especially interpreted. The experimental results show that, firstly, for a low pressure natural circulation system the two phase flow is unstable in most of inlet subcooling conditions, the two phase stable flow can only be reached at very low inlet subcooling; secondly, at high inlet subcooling the flow instability is dominated by subcooled boiling in the heated section, and at middle inlet subcooling is dominated by void flashing in the adiabatic long riser; thirdly, in two phase stable flow region the condition for boiling out of the core, namely, single phase flow in the heated section, two phase flow in the riser due to vapor flashing, can be realized. The experimental results are very important for the design and accident analysis of the vessel and swimming pool type natural circulation nuclear heating reactor. (7 refs., 10 figs., 1 tab.)

  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. Current limitation in low pressure mercury arcs

    International Nuclear Information System (INIS)

    Torven, S.; Babic, M.

    1976-06-01

    When the electric current in a low pressure arc with a long positive column is increased sufficiently, an electrostatic instability develops in the plasma which leads to formation of thin space charge layers across the column. The instability is investigated in a mercury plasma column kept axially homogeneous by a special technique. Values of some plasma parameters are measured at the instability threshold. It is found that the plasma is in a weakly ionized state in contrast to predictions by widely accepted current limitation theories. It is concluded that new types of theories are required to explain the observations. (Auth.)

  20. Thermal applications of low-pressure diamond

    International Nuclear Information System (INIS)

    Haubner, R.; Lux, B.

    1997-01-01

    During the last decade several applications of low-pressure diamond were developed. Main products are diamond heat-spreaders using its high thermal conductivity, diamond windows with their high transparency over a wide range of wavelengths and wear resistant tool coatings because of diamonds superhardness. A short description of the most efficient diamond deposition methods (microwave, DC-glow discharge, plasma-jet and arc discharge) is given. The production and applications of diamond layers with high thermal conductivity will be described. Problems of reproducibility of diamond deposition, the influence of impurities, the heat conductivity in electronic packages, reliability and economical mass production will be discussed. (author)

  1. 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

  2. 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.

  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. Development of High Efficiency and Low Emission Low Temperature Combustion Diesel Engine with Direct EGR Injection

    Science.gov (United States)

    Ho, R. J.; Kumaran, P.; Yusoff, M. Z.

    2016-03-01

    Focus on energy and environmental sustainability policy has put automotive research & development directed to developing high efficiency and low pollutant power train. Diffused flame controlled diesel combustion has reach its limitation and has driven R&D to explore other modes of combustions. Known effective mode of combustion to reduce emission are Low temperature combustion (LTC) and homogeneous charge combustion ignition by suppressing Nitrogen Oxide(NOx) and Particulate Matter (PM) formation. The key control to meet this requirement are chemical composition and distribution of fuel and gas during a combustion process. Most research to accomplish this goal is done by manipulating injected mass flow rate and varying indirect EGR through intake manifold. This research paper shows viable alternative direct combustion control via co-axial direct EGR injection with fuel injection process. A simulation study with OpenFOAM is conducted by varying EGR injection velocity and direct EGR injector diameter performed with under two conditions with non-combustion and combustion. n-heptane (C7H16) is used as surrogate fuel together with 57 species 290 semi-detailed chemical kinetic model developed by Chalmers University is used for combustion simulation. Simulation result indicates viability of co-axial EGR injection as a method for low temperature combustion control.

  5. An RF driven H- source and a low energy beam injection system for RFQ operation

    International Nuclear Information System (INIS)

    Leung, K.N.; Bachman, D.A.; Chan, C.F.; McDonald, D.S.

    1992-01-01

    An RF driven H - source has been developed at LBL for use in the Superconducting Super Collider (SSC). To date, an H - current of ∼40 mA can be obtained from a 5.6-cm-diam aperture with the source operated at a pressure of about 12 m Torr and 50 kW of RF power. In order to match the accelerated H - beam into the SSC RFQ, a low-energy H - injection system has been designed. This injector produces an outgoing H - beam free of electron contamination, with small radius, large convergent angle and small projectional emittance

  6. 49 CFR 192.623 - Maximum and minimum allowable operating pressure; Low-pressure distribution systems.

    Science.gov (United States)

    2010-10-01

    ... distribution systems. (a) No person may operate a low-pressure distribution system at a pressure high enough to...) No person may operate a low pressure distribution system at a pressure lower than the minimum... 49 Transportation 3 2010-10-01 2010-10-01 false Maximum and minimum allowable operating pressure...

  7. Aschroft Pressure Switch - Monitor for Low SCHe Supply Bottle Pressure

    International Nuclear Information System (INIS)

    VAN KATWIJK, C.

    2000-01-01

    These pressure switches are located in the SCHe helium supply lines at the pressure bottles and upstream of the PRV. The switches monitor the SCHe supply bottle pressure and are set to alarm at 2200 psig. There is one switch for each SCHe supply (4). Electronic output signal is NON-SAFETY (GS)

  8. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    Abass, O. A. M.

    2014-11-01

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  9. 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.

  10. 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.

  11. 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...

  12. 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

  13. Influence of injection pressures till to 1,000 bar on the carburetion in a spark ignition engine with direct injection; Einfluss von Einspritzdruecken bis 1000 bar auf die Gemischbildung in einem Ottomotor mit Direkteinspritzung

    Energy Technology Data Exchange (ETDEWEB)

    Buri, Stefan; Schumann, Florian; Kubach, Heiko; Spicher, Ulrich [Karlsruher Institut fuer Technologie (KIT) (DE). Inst. fuer Kolbenmaschinen (IFKM); Kneifel, Alexander [MTU Friedrichshafen GmbH (Germany)

    2011-07-01

    This paper presents the results of optical investigations of the impact of injection pressures of up to 1000 bar on mixture formation in a spray-guided direct injection engine. The maximum load in stratified operation of an engine with such a spray-guided combustion system is limited by the achievable quality of the mixture. In particular, when using multi hole injectors, the limit of stratified operation is reached rather early, due to comparatively low flow rates and thus insufficient stratification. One measure to increase the flow rate is to increase the injection pressure. The goal of this measure is to generate a more compact stratification, leading to combustion at richer air fuel ratios. This enables reductions of burning duration, hydrocarbon- and particulate emissions. The fundamental impact of increasing the injection pressure from 200 up to 1000 bar on mixture formation was investigated by using LIF- and Mie-scattering in a pressure chamber. Following that, the mixture formation was investigated under real conditions in a single cylinder engine by visualizing the injection process using Mie-scattering. Finally the results of engine operation are compared with those from the pressure chamber. (orig.)

  14. Concordant pressure paresthesia during interlaminar lumbar epidural steroid injections correlates with pain relief in patients with unilateral radicular pain.

    Science.gov (United States)

    Candido, Kenneth D; Rana, Maunak V; Sauer, Ruben; Chupatanakul, Lalida; Tharian, Antony; Vasic, Vladimir; Knezevic, Nebojsa Nick

    2013-01-01

    Transforaminal and interlaminar epidural steroid injections are commonly used interventional pain management procedures in the treatment of radicular low back pain. Even though several studies have shown that transforaminal injections provide enhanced short-term outcomes in patients with radicular and low back pain, they have also been associated with a higher incidence of unintentional intravascular injection and often dire consequences than have interlaminar injections. We compared 2 different approaches, midline and lateral parasagittal, of lumbar interlaminar epidural steroid injection (LESI) in patients with unilateral lumbosacral radiculopathic pain. We also tested the role of concordant pressure paresthesia occurring during LESI as a prognostic factor in determining the efficacy of LESI. Prospective, randomized, blinded study. Pain management center, part of a teaching-community hospital in a major metropolitan US city. After Institutional Review Board approval, 106 patients undergoing LESI for radicular low back pain were randomly assigned to one of 2 groups (53 patients each) based on approach: midline interlaminar (MIL) and lateral parasagittal interlaminar (PIL). Patients were asked to grade any pressure paresthesia as occurring ipsilaterally or contralaterally to their "usual and customary pain," or in a distribution atypical of their daily pain. Other variables such as: the Oswestry Disability Index questionnaire, pain scores at rest and during movement, use of pain medications, etc. were recorded 20 minutes before the procedure, and on days 1, 7, 14, 21, 28, 60, 120, 180 and 365 after the injection. Results of this study showed statistically and clinically significant pain relief in patients undergoing LESI by both the MIL and PIL approaches. Patients receiving LESI using the lateral parasagittal approach had statistically and clinically longer pain relief then patients receiving LESI via a midline approach. They also had slightly better quality of

  15. Influence of contrast media viscosity and temperature on injection pressure in computed tomographic angiography: a phantom study.

    Science.gov (United States)

    Kok, Madeleine; Mihl, Casper; Mingels, Alma A; Kietselaer, Bas L; Mühlenbruch, Georg; Seehofnerova, Anna; Wildberger, Joachim E; Das, Marco

    2014-04-01

    Iodinated contrast media (CM) in computed tomographic angiography is characterized by its concentration and, consecutively, by its viscosity. Viscosity itself is directly influenced by temperature, which will furthermore affect injection pressure. Therefore, the purposes of this study were to systematically evaluate the viscosity of different CM at different temperatures and to assess their impact on injection pressure in a circulation phantom. Initially, viscosity of different contrast media concentrations (240, 300, 370, and 400 mgI/mL) was measured at different temperatures (20°C-40°C) with a commercially available viscosimeter. In the next step, a circulation phantom with physical conditions was used. Contrast media were prepared at different temperatures (20°C, 30°C, 37°C) and injected through a standard 18-gauge needle. All other relevant parameters were kept constant (iodine delivery rate, 1.9 g I/s; total amount of iodine, 15 g I). Peak flow rate (in milliliter per second) and injection pressure (psi) were monitored. Differences in significance were tested using the Kruskal-Wallis test (Statistical Package for the Social Sciences). Viscosities for iodinated CM of 240, 300, 370, and 400 mg I/mL at 20°C were 5.1, 9.1, 21.2, and 28.8 mPa.s, respectively, whereas, at 40°C, these were substantially lower (2.8, 4.4, 8.7, and 11.2 mPa.s). In the circulation phantom, mean (SD) peak pressures for CM of 240 mg I/mL at 20°C, 30°C, and 37°C were 107 (1.5), 95 (0.6), and 92 (2.1) psi; for CM of 300 mg I/mL, 119 (1.5), 104 (0.6), and 100 (3.6) psi; for CM of 370 mg I/mL, 150 (0.6), 133 (4.4), and 120 (3.5) psi; and for CM of 400 mg I/mL, 169 (1.0), 140 (2.1), and 135 (2.9) psi, respectively, with all P values less than 0.05. Low concentration, low viscosity, and high temperatures of CM are beneficial in terms of injection pressure. This should also be considered for individually tailored contrast protocols in daily routine scanning.

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. Quality Management and Control of Low Pressure Cast Aluminum Alloy

    Science.gov (United States)

    Zhang, Dianxi; Zhang, Yanbo; Yang, Xiufan; Chen, Zhaosong; Jiang, Zelan

    2018-01-01

    This paper briefly reviews the history of low pressure casting and summarizes the major production processes of low pressure casting. It briefly introduces the quality management and control of low pressure cast aluminum alloy. The main processes include are: preparation of raw materials, Melting, refining, physical and chemical analysis, K-mode inspection, sand core, mold, heat treatment and so on.

  1. 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)

  2. The effect of low capacity injection systems on transient initiated loss of vessel water injection at Browns Ferry unit one

    International Nuclear Information System (INIS)

    Ott, L.T.

    1983-01-01

    Probabilistic risk assessment (PRA) analyses have indicated the transient initiated loss of vessel water injection (TQUV sequence) to be a dominant accident scenario for BWR plants. The PRA studies assumed the low capacity injection systems to be unimportant in severe accidents. The results of a Severe Accident Sequence Analysis (SASA) Program study have shown that these systems are capable of preventing or significantly delaying core damage in a TQUV sequence

  3. Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de inyeccion de agua de refrigeracion a baja presion (LPCI) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Delgado C, R. A.; Lopez S, E.; Chavez M, C., E-mail: renedelgado2015@hotmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

  4. 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

  5. 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

  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. A research program: The investigation of heat transfer and fluid flow at low pressure

    International Nuclear Information System (INIS)

    El-Genk, Mohamed S.; Philbin, Jeffrey S.; Foushee, Fabian C.

    1986-01-01

    This paper gives an overview of a multiyear joint research program being conducted at the University of New Mexico (UNM) with support from Sandia National Laboratories and GA Technologies. This research focuses on heat removal and fluid dynamics in flow regimes characterized by low pressure and low Reynolds number. The program was motivated by a desire to characterize and analyze cooling in a broad class of TRIGA-type reactors under: a) typical operating conditions, b) anticipated, new operating regimes, and c) postulated accident conditions. It has also provided experimental verification of analytical tools used in design analysis. The paper includes descriptions of the UNM thermal-hydraulics test facility and the experimental test sections. During the first two years experiments were conducted using single, electrically heated rod in water and air annuli. This configuration provides an observable and serviceable simulation of a fuel rod and its coolant channel. (author)

  8. Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

    OpenAIRE

    Lefèvre, Grégory; Živković, Ljiljana S.; Jaubertie, Anne

    2012-01-01

    In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspec...

  9. 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

  10. 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.

  11. Impacts of pellets injected from the low-field side on plasma in ITER

    International Nuclear Information System (INIS)

    Wisitsorasak, A.; Onjun, T.

    2011-01-01

    Impacts of pellets injected from the low-field side (LFS) on plasma in ITER are investigated using the 1.5D BALDUR integrated predictive modeling code. In these simulations, the pellet ablation is described using the neutral gas shielding (NGS) model. The pellet ablation model is coupled with the plasma core transport model, which is a combination of the MMM95 anomalous transport model and NCLASS neoclassical transport model. The boundary conditions are assumed to be at the top of the pedestal, in which the pedestal parameters are predicted using a pedestal model based on the theoretical-based pedestal width scaling (either magnetic and flow shear stabilization width scaling, or flow shear stabilization width scaling, or normalized poloidal pressure width scaling) and the infinite-n ballooning mode pressure gradient limit. These pedestal models depend sensitively on the density at the top of the pedestal, which can be strongly influenced by the injection of pellets. The combination of the MMM95 and NCLASS models, together with the pedestal and NGS models, is used to simulate the time evolution of the plasma current, ion and electron temperatures, and density profiles for ITER standard type-I ELMy H-mode discharges during the injection of LFS pellets. It is found that the injection of pellets results in a complicated plasma scenario, especially in the outer region of the plasma and the plasma conditions at the boundary in which the pellet has an impact on increasing the plasma edge density, but reducing the plasma edge temperature. The LFS pellet has a stronger impact on the edge as compared to the center. For fusion performance, the pellet can result in either enhancement or degradation, depending sensitively on the pellet parameters; such as the pellet size, pellet velocity, and pellet frequency. For example, when a series of deuterium pellets with a size of 0.5 cm, velocity of 1 km/s, and frequency of 2 Hz are injected into the ITER plasma from the LFS, the

  12. 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)

  13. Effects of molten material temperatures and coolant temperatures on vapor explosion

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; YUAN Minghao; HU Zhihua

    2007-01-01

    An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction(FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.

  14. 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

  15. 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

  16. Boron, arsenic and phosphorus dopant incorporation during low temperature low pressure silicon epitaxial growth

    International Nuclear Information System (INIS)

    Borland, J.O.; Thompson, T.; Tagle, V.; Benzing, W.

    1987-01-01

    Submicron silicon epitaxial structures with very abrupt epi/substrate transition widths have been realized through the use of low temperature silicon epitaxial growth techniques. At these low temperature and low pressure epitaxial growth conditions there is minimal, if any, dopant diffusion from the substrate into the epilayer during deposition. The reincorporation of autodoped dopant as well as the incorporation of intentional dopant can be a trade-off at low temperatures and low pressures. For advanced CMOS and Bi-CMOS technologies, five to six orders of magnitude change in concentration levels are desirable. In this investigation, all of the epitaxial depositions were carried out in an AMC-7810 epi-reactor with standard jets for a turbulent mixing system, and using a modified center inject configuration to achieve a single pass laminar flow system. To simulate the reincorporation of various autodoped dopant, the authors ran a controlled dopant flow of 100 sccm for each of the three dopants (boron, phosphorus and arsenic) to achieve the controlled background dopant level in the reactor gas stream

  17. Pressure suppression device

    International Nuclear Information System (INIS)

    Yoshida, Toyokazu.

    1976-01-01

    Purpose: To provide a pressure suppression device for a gas cooled reactor wherein the coolant is discharged in a reactor building by a loss-of-coolant accident or the like, the increase in the pressure and temperature is controlled and thermal energy of the discharged coolant of high temperature and high pressure can be absorbed. Constitution: A low heat source unit is provided at the upper part in an inner space of a reactor building provided around the reactor, and at the upper part of the low heat source unit a stirring fan for mixing gas within the building, and a low heat source circulating the low heat source through a pipe is connected to the low heat source unit. The low heat source unit is provided with the pipe arranged in a spiral shape at the upper part of the space of the unit, and a large number of fins are provided at the outer surface of the pipe for increasing the transmission area and improve the heat exchange. When the coolant of high temperature and high pressure has been lost in the building, the thermal energy of the coolant is absorbed by the low heat source unit. (Aizawa, K.)

  18. 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.

  19. 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.

  20. 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.

  1. Effects of CO2 injection and Kerogen Maturation on Low-Field Nuclear Magnetic Resonance Response

    Science.gov (United States)

    Prasad, M.; Livo, K.

    2017-12-01

    Low-field Nuclear Magnetic Resonance (NMR) is commonly used in petrophysical analysis of petroleum reservoir rocks. NMR experiments record the relaxation and polarization of in-situ hydrogen protons present in gaseous phases such as free-gas intervals and solution gas fluids, bulk fluid phases such as oil and aquifer intervals, and immovable fractions of kerogen and bitumen. Analysis of NMR relaxation spectra is performed to record how fluid composition, maturity, and viscosity change NMR experimental results. We present T1-T2 maps as thermal maturity of a water-saturated, sub-mature Woodford shale is increased at temperatures from 125 to 400 degrees Celsius. Experiments with applied fluid pressure in paraffinic mineral oil and DI water with varying fluid pH have been performed to mimic reservoir conditions in analysis of the relaxation of bulk fluid phases. We have recorded NMR spectra, T1-T2 maps, and fluid diffusion coefficients using a low-field (2 MHz) MagritekTM NMR. CO2 was injected at a pressure of 900 psi in an in house developed NMR pressure vessel made of torlon plastic. Observable 2D NMR shifts in immature kerogen formations as thermal maturity is increased show generation of lighter oils with increased maturity. CO2 injection leads to a decrease in bulk fluid relaxation time that is attributed to viscosity modification with gas presence. pH variation with increased CO2 presence were shown to not effect NMR spectra. From this, fluid properties have been shown to greatly affect NMR readings and must be taken into account for more accurate NMR reservoir characterization.

  2. 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.

  3. Organic coolant in Winnipeg riverbed sediments

    International Nuclear Information System (INIS)

    Guthrie, J.E.; Acres, O.E.

    1979-03-01

    Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

  4. 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)

  5. Vibration analysis of the Golfech 2 safety injection system

    International Nuclear Information System (INIS)

    Morilhat, P.

    1993-01-01

    The main function of the safety injection system in a PWR plant is to ensure cooling of fuel elements in the event of a loss of coolant accident. The multistage centrifugal pump mounted-on this system induces pressure fluctuations, resulting in dynamic loads on piping. In certain plant units, these loads have caused cracking in the nozzles connected to the safety injection system, whereas in others, no damage has been observed. In order to understand the differences in dynamic behavior observed from one site to another, tests were performed on a real safety injection system, that of Golfech-2. They enabled determination of the modal characteristics of the system and identification of the hydro-acoustic source of the low head safety injection pump. They also enabled assessment of the pressure fluctuation levels in the pump suction and discharge areas as well as the vibratory response of the system when operating under partial and nominal flow conditions. Finally, these test results were used to estimate fatigue damage in the safety injection system. The experimental results will later be used to validate the model of the system undertaken with the piping design code CIRCUS and define the boundary conditions to be taken into account. (author). 6 figs., 2 refs

  6. Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

    International Nuclear Information System (INIS)

    Luo Bangqi

    2012-01-01

    The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

  7. Short-term intraocular pressure changes after intravitreal injection of bevacizumab in diabetic retinopathy patients

    Directory of Open Access Journals (Sweden)

    Farhood QK

    2014-03-01

    Full Text Available Qasim Kadhim Farhood,1 Sinan Mohammad Twfeeq21College of Medicine, Babylon University, Babylon; 2Al-Jumhori teaching hospital, Mosul, IraqBackground: This study examined the changes in short-term intraocular pressure (IOP in a prospective series of patients undergoing intravitreal bevacizumab injection. The aim was to evaluate the frequency and predictive factors related to intraocular pressure (IOP elevation in patients receiving intravitreal bevacizumab.Patients and methods: This study included 52 patients with diabetic retinopathy between 28 to 75 years of age with a mean age of 51 years; 30 (58% were females, and 22 (42% were males. All patients received bevacizumab (1.25 mg/0.05 mL injected intravitreally in a standard fashion between May 2012 to February 2013 in the AL-Jumhoury teaching hospital. IOP was measured at baseline, 5, 10, and 30 minutes after injection using Goldman applanation tonometry.Statistics: Data were analyzed using the SPSS v.12.0 for windows. Basic, demographic, and clinical data were analyzed using means, proportions, and appropriate 95% confidence intervals (CIs.Results: Most patients (85% were diagnosed with proliferative diabetic retinopathy, while 15% presented with diabetic macular edema. The mean IOP values at baseline, 5, 10, and 30 minutes after injection were 14.0 mmHg (95% CI 13.4–14.7, 36.1 mmHg (95% CI 33.5–38.6, 25.7 mmHg (95% CI 23.8–27.5, and 15.5 mmHg (95% CI 12.4–16.51, respectively. Regression analysis showed a trend toward phakic patients having higher IOP at 30 minutes.Conclusion: Intravitreal injection of bevacizumab is safe with respect to short-term IOP changes, as almost all IOP returned to a safe range (<25 mmHg within 30 minutes. Elevated IOP 30 minutes after injection only occurs rarely, so routine prophylactic use of anti-glaucoma medication is not indicated.Keywords: intravitreal injection, bevacizumab, intraocular pressure, Goldmann applanation tonometry

  8. 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

  9. Trace organics in AGR coolants

    International Nuclear Information System (INIS)

    Smith, R.; Green, L.O.; Johnson, P.A.V.

    1980-01-01

    Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

  10. Investigation of low emission combustors using hydrogen lean direct injection

    Directory of Open Access Journals (Sweden)

    Robert ISAC

    2011-09-01

    Full Text Available One of the key technology challenges for the use of hydrogen in gas turbine engines is the performance of the combustion system, in particular the fuel injectors. Tests were conducted to measure the nitrogen oxide (NOx emissions and combustion performance at inlet conditions of 588 to 811 K, 0.4 to 1.4 MPa, and equivalence ratios up to 0.48. All the injectors were based on Lean Direct Injection (LDI technology with multiple injection points and quick mixing. One challenge to hydrogen-based premixing combustion systems is flashback since hydrogen has a reaction rate over 7 times that of Jet-A.

  11. Formation and sustainment of a low aspect ratio tokamak by a series of plasma injections

    International Nuclear Information System (INIS)

    Shimamura, Shin; Taniguchi, Makoto; Takahashi, Tsutomu; Nogi, Yasuyuki

    1995-01-01

    A low aspect ratio tokamak plasma was generated and sustained by injecting a series of plasmas from a magnetized coaxial gun into a flux conserver with toroidal field. The magnetized coaxial gun was supplied by an oscillating current with a d.c. component. The first few current pulses injected plasma and helicity into the flux conserver. This pulse helicity injection method worked effectively to maintain the low aspect ratio tokamak. 8 refs., 5 figs

  12. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.; Hollies, B.

    1986-08-01

    Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

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

    Science.gov (United States)

    Marra, John J

    2014-04-01

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

  14. 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

  15. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  16. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  17. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  18. Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

    Science.gov (United States)

    Feddema, Rick

    fuels. Optical patternation data and line of sight laser diffraction data show that there is significant difference between jet fuels. Particularly at low fuel injection pressures (0.345 MPa) and cold temperatures (-40 C), the patternation data shows that the total surface area in the spray at 38.1 mm from the pressure swirl injector for the JP-10 fuel type is one-sixth the amount of the JP-8. Finally, this study compares the atomizer performance of a pressure swirl nozzle to a hybrid air blast nozzle. The total surface area for both the hybrid air blast nozzle and the pressure swirl nozzle show a similar decline in atomization performance at low fuel injection pressures and cold temperatures. However, the optical patternator radial profile data and the line of sight laser diffraction data show that the droplet size and spray distribution data are less affected by injection conditions and fuel type in the hybrid air blast nozzle, than they are in the pressure swirl nozzle. One explanation is that the aerodynamic forces associated with the swirler on the hybrid air blast nozzle control the distribution droplets in the spray. This is in contrast to the pressure swirl nozzle droplet distribution that is controlled by internal geometry and droplet ballistics.

  19. Effect of injection pressure on performance, emission, and combustion characteristics of diesel-acetylene-fuelled single cylinder stationary CI engine.

    Science.gov (United States)

    Srivastava, Anmesh Kumar; Soni, Shyam Lal; Sharma, Dilip; Jain, Narayan Lal

    2018-03-01

    In this paper, the effect of injection pressure on the performance, emission, and combustion characteristics of a diesel-acetylene fuelled single cylinder, four-stroke, direct injection (DI) diesel engine with a rated power of 3.5 kW at a rated speed of 1500 rpm was studied. Experiments were performed in dual-fuel mode at four different injection pressures of 180, 190, 200, and 210 bar with a flow rate of 120 LPH of acetylene and results were compared with that of baseline diesel operation. Experimental results showed that highest brake thermal efficiency of 27.57% was achieved at injection pressure of 200 bar for diesel-acetylene dual-fuel mode which was much higher than 23.32% obtained for baseline diesel. Carbon monoxide, hydrocarbon, and smoke emissions were also measured and found to be lower, while the NO x emissions were higher at 200 bar in dual fuel mode as compared to those in other injection pressures in dual fuel mode and also for baseline diesel mode. Peak cylinder pressure, net heat release rate, and rate of pressure rise were also calculated and were higher at 200 bar injection pressure in dual fuel mode.

  20. Properties of low-pressure drift chambers

    International Nuclear Information System (INIS)

    Breskin, A.; Trautner, N.

    1976-01-01

    Drift chambers operated with methylal vapour or ethylene at pressures in the range of 10-110 torr are described. A systematic study of position resolution, pulse height and rise time shows that especially for ethylene they are strongly influenced by electron diffusion. Intrinsic position resolution was found to be at least as good as found at atmospheric pressure. A relative pulse height resolution of 10% was obtained with 5.5 MeV alpha-particles. A simple mathematical model which can describe the processes in the drift chamber is presented. (Auth.)

  1. 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.

  2. Pressurized air injection in an axial hydro-turbine model for the mitigation of tip leakage cavitation

    Science.gov (United States)

    Rivetti, A.; Angulo, M.; Lucino, C.; Liscia, S.

    2015-12-01

    Tip leakage vortex cavitation in axial hydro-turbines may cause erosion, noise and vibration. Damage due to cavitation can be found at the tip of the runner blades on the low pressure side and the discharge ring. In some cases, the erosion follows an oscillatory pattern that is related to the number of guide vanes. That might suggest that a relationship exists between the flow through the guide vanes and the tip vortex cavitating core that induces this kind of erosion. On the other hand, it is known that air injection has a beneficial effect on reducing the damage by cavitation. In this paper, a methodology to identify the interaction between guide vanes and tip vortex cavitation is presented and the effect of air injection in reducing this particular kind of erosion was studied over a range of operating conditions on a Kaplan scale model. It was found that air injection, at the expense of slightly reducing the efficiency of the turbine, mitigates the erosive potential of tip leakage cavitation, attenuates the interaction between the flow through the guide vanes and the tip vortex and decreases the level of vibration of the structural components.

  3. Study on enhanced lymphatic tracing of isosulfan blue injection by influence of osmotic pressure on lymphatic exposure.

    Science.gov (United States)

    Ye, Tiantian; He, Rui; Wu, Yue; Shang, Lei; Wang, Shujun

    2018-04-01

    Isosulfan blue (IB) is being used as a lymphatic tracer has been approved by the FDA in 1981. This study aimed at improving lymphatic exposure of IB injection by osmotic pressure regulation to achieve step-by step lymphatic tracing. First, IB injection with appropriate osmotic pressure, stability, and suitable pH was prepared. Next, the lymphatic tracing ability of different osmotic pressure was studied to determine the blue-stained state of IB in three-level lymph nodes after subcutaneous administration. Furthermore, pharmacokinetics of lymphatic drainage, lymph node uptake, and plasma concentration was investigate to explore the improving law of the lymphatic tracing by osmotic pressure, and combined with tissue irritation to determine the optimal osmotic pressure. At last, the tissue distribution in mice of IB injection which had the property of optimal osmotic pressure was investigated. The results showed that increasing osmotic pressure could significantly reduce injection site retention and increase IB concentration of lymph node. The lymph nodes could be obviously blue-stained by IB injection which had 938 mmol/kg osmotic pressure and would not cause inflammatory reaction and blood exposure. The tissue distribution study suggested that IB injection which had 938 mmol/kg osmotic pressure was mainly distributed into gallbladder and duodenum that verified the reports that 90% IB was excreted through the feces through biliary excretion. In conclusion, this study provides the basic study to improve lymphatic exposure of IB injection by regulate the osmotic pressure and have the potential to be the helpful guidance for the elective lymph node dissection.

  4. Low pressure gage type VM-01

    International Nuclear Information System (INIS)

    Brandea, I.; Curuia, M.; Culcer, M.

    2000-01-01

    High vacuum systems became an important element of many applied technologies, from gas analysers to rocket engines. An intelligent apparatus for pressure measurement in the range of 10 -3 - 10 -8 mbar, with incorporated INTEL 80C51 microcontroller is presented. Based on a Bayard-Alpert hot cathode gage, equally developed in our institute, the pressure gage allows the displaying of different operation parameters and also of the error codes for different kinds of malfunctioning, as for instance missing of grid voltage, grid-collector breakdown, pressure increasing above 10 -3 mbar. Its operation is based on a microcontroller assembly language program especially worked out and introduced in the central units EPROM memory. The gage characteristics for different gases are also introduced in an EPROM memory, and the type of the gas is selected by the operator from the front panel. One can select also from the front panel the pressure unit (mbar, torr, Pa). If a remote control is necessary, this can be done by means of a PC, with a program written in the LabVIEW graphical programming language. The pressure gage was tested and calibrated in relation with an EDWARDS vacuum measuring system and provided a good accuracy (better than 25%). Its field of application is both laboratory and industrial measurements. Its main features are: - supply voltage, 220 V ac / 50 Hz; - power consumption, 30 W; - gage's grid supply voltage, 160 V; grid current, 2 mA / p = 10 -5 ...10 -8 mbar and 0.2 mA / p = 10 -3 ...10 -5 mbar; - cathode heating current, max. 3 A; - measuring range, 10 -3 ...10 -8 mbar; - error of measurement, ±35%; - remote control, according to the RS232 standard; - size, 320 x 200 x 100 mm; - weight, 3.5 Kg. (authors)

  5. Shock tubes: compressions in the low pressure chamber

    International Nuclear Information System (INIS)

    Schins, H.; Giuliani, S.

    1986-01-01

    The gas shock tube used in these experiments consists of a low pressure chamber and a high pressure chamber, divided by a metal-diaphragm-to-rupture. In contrast to the shock mode of operation, where incident and reflected shocks in the low pressure chamber are studied which occur within 3.5 ms, in this work the compression mode of operation was studied, whose maxima occur (in the low pressure chamber) about 9 ms after rupture. Theoretical analysis was done with the finite element computer code EURDYN-1M, where the computation was carried out to 30 ms

  6. [Injection Pressure Evaluation of the New Venous Catheter with Side Holes for Contrast-enhanced CT/MRI].

    Science.gov (United States)

    Fukuda, Junya; Arai, Keisuke; Miyazawa, Hitomi; Kobayashi, Kyouko; Nakamura, Junpei; Suto, Takayuki; Tsushima, Yoshito

    2018-01-01

    The simulation study was conducted for the new venous catheter with side holes of contrast-enhanced computed tomography (CT) and magnetic resonance imaging (MRI) to evaluate the infusion pressure on four contrast media and several injection speeds. All infusion pressure of the new venous catheter with side holes were less than 15 kg/cm 2 as limitation of extension tube and also reduced the infusion pressure by 15% at the maximum compared to the catheter with single hole. The results suggest that the new venous catheter with side holes can reduce the infusion pressure by power injection of contrast-enhanced CT and MRI.

  7. Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

    International Nuclear Information System (INIS)

    Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

    1983-01-01

    The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

  8. MODEL TESTING OF LOW PRESSURE HYDRAULIC TURBINE WITH HIGHER EFFICIENCY

    Directory of Open Access Journals (Sweden)

    V. K. Nedbalsky

    2007-01-01

    Full Text Available A design of low pressure turbine has been developed and it is covered by an invention patent and a useful model patent. Testing of the hydraulic turbine model has been carried out when it was installed on a vertical shaft. The efficiency was equal to 76–78 % that exceeds efficiency of the known low pressure blade turbines. 

  9. The sodium coolant

    International Nuclear Information System (INIS)

    Rodriguez, G.

    2004-01-01

    The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

  10. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  11. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 06-06-2016 2. REPORT TYPE Interim Report 3. DATES COVERED ... Corrosion Testing of Traditional and Extended Life Coolants 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Hansen, Gregory A. T...providing vehicle specific coolants. Several laboratory corrosion tests were performed according to ASTM D1384 and D2570, but with a 2.5x extended time

  12. Sector Tests of a Low-NO(sub x), Lean, Direct- Injection, Multipoint Integrated Module Combustor Concept Conducted

    Science.gov (United States)

    Tacina, Robert R.; Wey, Chang-Lie; Laing, Peter; Mansour, Adel

    2002-01-01

    The low-emissions combustor development described is directed toward advanced high pressure aircraft gas-turbine applications. The emphasis of this research is to reduce nitrogen oxides (NOx) at high-power conditions and to maintain carbon monoxide and unburned hydrocarbons at their current low levels at low power conditions. Low-NOx combustors can be classified into rich-burn and lean-burn concepts. Lean-burn combustors can be further classified into lean-premixed-prevaporized (LPP) and lean direct injection (LDI) concepts. In both concepts, all the combustor air, except for liner cooling flow, enters through the combustor dome so that the combustion occurs at the lowest possible flame temperature. The LPP concept has been shown to have the lowest NOx emissions, but for advanced high-pressure-ratio engines, the possibility of autoignition or flashback precludes its use. LDI differs from LPP in that the fuel is injected directly into the flame zone, and thus, it does not have the potential for autoignition or flashback and should have greater stability. However, since it is not premixed and prevaporized, good atomization is necessary and the fuel must be mixed quickly and uniformly so that flame temperatures are low and NOx formation levels are comparable to those of LPP. The LDI concept described is a multipoint fuel injection/multiburning zone concept. Each of the multiple fuel injectors has an air swirler associated with it to provide quick mixing and a small recirculation zone for burning. The multipoint fuel injection provides quick, uniform mixing and the small multiburning zones provide for reduced burning residence time, resulting in low NOx formation. An integrated-module approach was used for the construction where chemically etched laminates, diffusion bonded together, combine the fuel injectors, air swirlers, and fuel manifold into a single element. The multipoint concept combustor was demonstrated in a 15 sector test. The configuration tested had 36

  13. Seed disinfection effect of atmospheric pressure plasma and low pressure plasma on Rhizoctonia solani.

    Science.gov (United States)

    Nishioka, Terumi; Takai, Yuichiro; Kawaradani, Mitsuo; Okada, Kiyotsugu; Tanimoto, Hideo; Misawa, Tatsuya; Kusakari, Shinichi

    2014-01-01

    Gas plasma generated and applied under two different systems, atmospheric pressure plasma and low pressure plasma, was used to investigate the inactivation efficacy on the seedborne pathogenic fungus, Rhizoctonia solani, which had been artificially introduced to brassicaceous seeds. Treatment with atmospheric plasma for 10 min markedly reduced the R. solani survival rate from 100% to 3% but delayed seed germination. The low pressure plasma treatment reduced the fungal survival rate from 83% to 1.7% after 10 min and the inactivation effect was dependent on the treatment time. The seed germination rate after treatment with the low pressure plasma was not significantly different from that of untreated seeds. The air temperature around the seeds in the low pressure system was lower than that of the atmospheric system. These results suggested that gas plasma treatment under low pressure could be effective in disinfecting the seeds without damaging them.

  14. Low-Cost Fiber Optic Pressure Sensor

    Science.gov (United States)

    Sheem, Sang K.

    2003-07-22

    The size and cost of fabricating fiber optic pressure sensors is reduced by fabricating the membrane of the sensor in a non-planar shape. The design of the sensors may be made in such a way that the non-planar membrane becomes a part of an air-tight cavity, so as to make the membrane resilient due to the air-cushion effect of the air-tight cavity. Such non-planar membranes are easier to make and attach.

  15. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  16. Ozone kinetics in low-pressure discharges

    Science.gov (United States)

    Guerra, Vasco; Marinov, Daniil; Guaitella, Olivier; Rousseau, Antoine

    2012-10-01

    Ozone kinetics is quite well established at atmospheric pressure, due to the importance of ozone in atmospheric chemistry and to the development of industrial ozone reactors. However, as the pressure is decreased and the dominant three-body reactions lose importance, the main mechanisms involved in the creation and destruction of ozone are still surrounded by important uncertainties. In this work we develop a self-consistent model for a pulsed discharge and its afterglow operating in a Pyrex reactor with inner radius 1 cm, at pressures in the range 1-5 Torr and discharge currents of 40-120 mA. The model couples the electron Boltzmann equation with a system of equations for the time evolution of the heavy particles. The calculations are compared with time-dependent measurements of ozone and atomic oxygen. Parametric studies are performed in order to clarify the role of vibrationally excited ozone in the overall kinetics and to establish the conditions where ozone production on the surface may become important. It is shown that vibrationally excited ozone does play a significant role, by increasing the time constants of ozone formation. Moreover, an upper limit for the ozone formation at the wall in these conditions is set at 10(-4).

  17. Method and apparatus for suppressing water-solid overpressurization of coolant in nuclear reactor power apparatus

    International Nuclear Information System (INIS)

    Aanstad, O.J.; Sklencar, A.M.

    1983-01-01

    A reactor-coolant relief valve is opened for increase in mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). (author)

  18. 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.

  19. Low-frequency current drive and helicity injection

    International Nuclear Information System (INIS)

    Chan, V.S.; Miller, R.L.; Ohkawa, T.

    1990-01-01

    For ω much-lt Ω i , where Ω i is the ion cyclotron frequency, circularly polarized waves can drive current far exceeding the current resulting from linearly polarized waves. Further, the efficiency can be independent of plasma density. In some cases, this circular polarization may be interpreted in terms of helicity injection. For tokamak applications, where the wavenumber in the toroidal direction is a real quantity, wave helicity is injected only with finite E z waves, where z is the direction of the static magnetic field. The Alfven waves are possible current drive candidates but, in the cylindrical model considered, the compressional wave is weakly damped because E z =0, while the shear Alfven wave is totally absorbed at the surface because of finite E z . A mixture of the two modes is shown to drive an oscillatory surface current even though the efficiency is high and independent of density. A more promising current drive candidate is a fast wave that propagates to the plasma interior and is damped by the minority cyclotron resonance. Near the minority mode conversion region, the fast wave is left-handed circularly polarized and it has a small but finite E z component at high electron temperatures. The current drive efficiency, although not as high as that of the Alfven wave, is still good and independent of density, making it attractive for fusion reactors

  20. 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)

  1. The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

    International Nuclear Information System (INIS)

    Hermansky, P.; Krajcovic, M.

    2012-01-01

    The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)

  2. Experimental Studies of Diestrol-Micro Emulsion Fuel in a Direct Injection Compression Ignition Engine under Varying Injection Pressures and Timings

    Science.gov (United States)

    Kannan, Gopal Radhakrishnan

    2018-02-01

    The research work on biodiesel becomes more attractive in the context of limited availability of petroleum fuels and rapid increase of harmful emissions from diesel engine using conventional fossil fuels. The present investigation has dealt with the influence of biodiesel-diesel-ethanol (diestrol) water micro emulsion fuel (B60D20E20M) on the performance, emission and combustion characteristics of a diesel engine under different injection pressure and timing. The results revealed that the maximum brake thermal efficiency of 32.4% was observed at an injection pressure of 260 bar and injection timing of 25.5°bTDC. In comparison with diesel, micro emulsion fuel showed reduction in carbon monoxide (CO) and total hydrocarbon (THC) by 40 and 24%, respectively. Further, micro emulsion fuel decreased nitric oxide (NO) emission and smoke emission by 7 and 20.7%, while the carbon dioxide (CO2) emission is similar to that of diesel.

  3. Hydropyrolysis of coal at very low pressure

    Energy Technology Data Exchange (ETDEWEB)

    Bi, J.; Kamo, T.; Kodera, Y.; Yamaguchi, H.; Sato, Y. [National Institute for Resources and Environment, Tsukuba-city (Japan). Energy Resources Department

    1998-07-01

    In the present study, Taiheiyo coal was heated under a pressure of 0.6 Torr and a temperature range of 20-800{degree}C in the presence and absence of hydrogen atoms. The yielded char, gas and liquid were measured and analyzed by TG-MS and GC-MS respectively. The result shows that the conversion of coal with hydrogen atoms is higher than that without, and more gas and liquid products are obtained in the presence of hydrogen atoms. 3 refs., 5 figs.

  4. 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.

  5. 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.

  6. 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.

  7. Contact condensation effects in the main coolant pipe

    International Nuclear Information System (INIS)

    Haefner, W.; Fischer, K.

    1990-01-01

    Contact condensation effects may occur in a pressurized water reactor (PWR) after a loss of coolant accident (LOCA) when emergency core cooling (ECC) water is injected contact with escaping steam which is generated within the core. The condensation which takes place may cause a sudden depressurization leading to the formation of water slugs. The interaction between the transient condensation and the inertia of the flow may also result in large amplitude flow and pressure oscillations. These contact condensation effects are of great importance for the mass flow distribution and the coolant water supply to the reactor core. To examine those complex processes, large computer codes are necessary. The development and verification of analytical models requires greatly simplified flow boundary conditions from experiments and a sufficiently large base of experimental data. Separate models have been developed for interfacial exchange of mass, momentum and energy with respect to the associated flow regime. Therefore, an adequate description of the condensation process requires the modeling of two different topics: the prediction of the flow regime and the calculation of the interfacial exchange. (author)

  8. Gas Injection And Fast Pressure-Rise Measurements For The Linac4 H− Source

    CERN Document Server

    Mahner, E; Lettry, J; Mattei, S; O'Neil, M; Neupert, H; Pasquino, C; Schmitzer, C

    2013-01-01

    In the era of the Large Hadron Collider, the CERN injector complex comprising the 34 years old Linac2 with its primary proton source, is presently upgraded with a new linear accelerator for H− (Linac4). The design, construction, and test of volume production and cesiated RF-driven H− ion sources is presently ongoing with the final goal of producing an H− beam with 80 mA beam current, 45 keV beam energy, 500 s pulse length, and a repetition rate of 2 Hz. In order to have quantitative information of the hydrogen gas density at the moment of plasma ignition the dynamic vacuum properties of the plasma generator were studied experimentally. We describe the experimental setup and present fast pressure-rise measurements for different parameters of the gas injection system, such as gas species (H2, He, N2, Ar), piezo valve voltage pulse length (200 - 500 s), and injection pressure (400 - 2800 mbar). The obtained data are compared with a conductance model of the plasma generator.

  9. Efeito do processamento em misturas de alumina/ligantes orgânicos usadas na moldagem por injeção em baixa pressão Effect of processing variables in alumina/organic binders mixtures used in low-pressure injection molding

    Directory of Open Access Journals (Sweden)

    P . A. Ourique

    2013-03-01

    Full Text Available A moldagem por injeção em baixa pressão (MIBP é uma técnica que já vem sendo empregada na produção de peças cerâmicas com formas e geometrias complexas. A homogeneidade da mistura de ligantes orgânicos e pós cerâmicos é um fator determinante que deve ser controlado para minimizar a formação de imperfeições no processamento de feedstocks para MIBP. Defeitos típicos de processamento por MIBP, como bolhas de ar e aglomerados, geram gradientes de densidade nas misturas que, após conformação, possuem poucas possibilidades de remoção. Essas imperfeições comprometem o desempenho dos produtos obtidos por essa técnica. Este trabalho está focado na avaliação dessas heterogeneidades e como elas podem ser correlacionadas com a variação da densidade aparente e com o comportamento reológico dessas misturas. Para tanto, aluminas submicrométricas, como recebida e desaglomerada, foram adicionadas a uma mistura fundida de ligantes a base de parafinas, ceras e aditivos e processada em dois tipos diferentes de misturadores, com e sem o auxílio de vácuo. Foi observada a presença de aglomerados existentes na alumina como recebida, possivelmente gerados durante a etapa de calcinação. Também foi observado que o tipo de misturador e a aplicação ou não de vácuo durante a etapa final do processamento têm grande influência no tempo de mistura necessário para reduzir a viscosidade do feedstock para a injeção.The low-pressure injection molding (LPIM is a technique already being used in the production of ceramic parts with complex shapes and geometries. The homogeneity of the mixture of organic binder and ceramic powder is a determining factor which must be controlled to minimize defects formation while feedstock processing to LPIM. Typical defects of LPIM processing, such as air bubbles and agglomerates, generate density gradients in the mixtures, which, after shaping, have little possibility of removal. These imperfections

  10. Performance Analysis of Abrasive Waterjet Machining Process at Low Pressure

    Science.gov (United States)

    Murugan, M.; Gebremariam, MA; Hamedon, Z.; Azhari, A.

    2018-03-01

    Normally, a commercial waterjet cutting machine can generate water pressure up to 600 MPa. This range of pressure is used to machine a wide variety of materials. Hence, the price of waterjet cutting machine is expensive. Therefore, there is a need to develop a low cost waterjet machine in order to make the technology more accessible for the masses. Due to its low cost, such machines may only be able to generate water pressure at a much reduced rate. The present study attempts to investigate the performance of abrasive water jet machining process at low cutting pressure using self-developed low cost waterjet machine. It aims to study the feasibility of machining various materials at low pressure which later can aid in further development of an effective low cost water jet machine. A total of three different materials were machined at a low pressure of 34 MPa. The materials are mild steel, aluminium alloy 6061 and plastics Delrin®. Furthermore, a traverse rate was varied between 1 to 3 mm/min. The study on cutting performance at low pressure for different materials was conducted in terms of depth penetration, kerf taper ratio and surface roughness. It was found that all samples were able to be machined at low cutting pressure with varied qualities. Also, the depth of penetration decreases with an increase in the traverse rate. Meanwhile, the surface roughness and kerf taper ratio increase with an increase in the traverse rate. It can be concluded that a low cost waterjet machine with a much reduced rate of water pressure can be successfully used for machining certain materials with acceptable qualities.

  11. Non-equilbrium behavior of low-pressure plasma jets

    International Nuclear Information System (INIS)

    Chang, C.H.; Pfender, E.

    1989-01-01

    After establishing the basic equations, some sample calculations are presented to examine the thermodynamic state of the plasma from atmospheric to low pressures (80 mbar). These results indicate the validity of local thermodynamic equilibrium (LTE) at atmospheric pressure as well as strong deviations from LTE at lower pressures especially in terms of chemical equilibrium. Departures from kinetic equilibrium are not as severe as those from chemical equilibrium along the centerline of the jet. However, there are some departures from transitional equilibrium in the fringes of the jet. It is demonstrated that conventional methods based on the LTE assumption are not appropriate for describing low-pressure plasma jets

  12. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary

  13. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

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

    International Nuclear Information System (INIS)

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

    1977-03-01

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

  15. Experimental study an a low pressure solar still

    International Nuclear Information System (INIS)

    Sriram, V.; Kalidasa Murugavel, K.; Samuel Hansen, R.

    2013-01-01

    In this work, a low pressure, single basin double slope Solar Still was fabricated and tested for different depths and with different wick and porous materials. A vacuum pump was used to maintain low pressure inside the still. External condenser was used to condense the vapor with raw water as cooling fluid. The performance of the low pressure still was compared with conventional still. The total production of the still was increased by 88.66% when a light cotton cloth was used as wick material in the basin along with minimum depth of water. (authors)

  16. Experiments on the behaviour of thermite melt injected into sodium: Final report on the THINA test results

    International Nuclear Information System (INIS)

    Huber, F.; Kaiser, A.; Peppler, W.

    1994-01-01

    During hypothetical accidents of fast breeder reactors the core melts and part of the core material inventory is ejected into the upper coolant plenum. As a consequence, a fuel to coolant thermal interaction occurs between the melt and the sodium. A series of simulating experiments was carried out in KfK/IRS to improve the knowledge about the phenomenology of molten fuel/coolant interactions and to support theoretical work on the safety of fast breeder reactors. In the tests, a thermite melt of up to 3270 K is injected from below into a sodium pool the temperature of which is between 770 and 820 K. The masses of the melt and the sodium are about five and 150 kg, respectively. Thermal interactions have been observed to occur as a sequence of small local pressure events mainly during the melt injection. Large-scale vapour explosions have not been observed. Generally, the conversion ratios of thermal to mechanical energy have been low. (author)

  17. Low pressure hugoniot cusp in polymeric materials

    Science.gov (United States)

    Sheffield, S. A.; Bloomquist, D. D.

    1982-04-01

    It has previously been shown that polymethyl methacrylate (PMMA) exhibits a cusp in the shock Hugoniot at about 2.0 GPa which corresponds with the beginning of shock-induced polarization and the beginning of an exothermic reaction measured in thermocouple and resistivity gauge temperature studies. We now report results we have recently obtained from an ongoing study which indicate that other polymers have similar behavior at about the same pressure. Quartz gauge impact experiments have been performed using polypyro-ellitimide (Vespel) and polysulfone impactors to obtain Hugoniot information and the stress history at the impact plane. In the case of Vespel a slight Hugoniot cusp was observed at about 1.8 GPa which coincides with the start of shock-induced polarization. Polysulfone does not appear to have a cusp but does show stress relaxation at the impact plane beginning at about 1.8 GPa, again coinciding with the start of shock-induced polarization. It has been suggested earlier that the abnormal behavior in PMMA is the result of a shock-induced chemical reaction. This new information suggests that a stress of about 2 GPa is a threshold for shock-induced chemical reaction in several polymers.

  18. Low Power and High Sensitivity MOSFET-Based Pressure Sensor

    International Nuclear Information System (INIS)

    Zhang Zhao-Hua; Ren Tian-Ling; Zhang Yan-Hong; Han Rui-Rui; Liu Li-Tian

    2012-01-01

    Based on the metal-oxide-semiconductor field effect transistor (MOSFET) stress sensitive phenomenon, a low power MOSFET pressure sensor is proposed. Compared with the traditional piezoresistive pressure sensor, the present pressure sensor displays high performances on sensitivity and power consumption. The sensitivity of the MOSFET sensor is raised by 87%, meanwhile the power consumption is decreased by 20%. (cross-disciplinary physics and related areas of science and technology)

  19. Low-level wind response to mesoscale pressure systems

    Science.gov (United States)

    Garratt, J. R.; Physick, W. L.

    1983-09-01

    Observations are presented which show a strong correlation between low-level wind behaviour (e.g., rotation near the surface) and the passage of mesoscale pressure systems. The latter are associated with frontal transition zones, are dominated by a pressure-jump line and a mesoscale high pressure area, and produce locally large horizontal pressure gradients. The wind observations are simulated by specifying a time sequence of perturbation pressure gradient and subsequently solving the vertically-integrated momentum equations with appropriate initial conditions. Very good agreement is found between observed and calculated winds; in particular, (i) a 360 ° rotation in wind on passage of the mesoscale high; (ii) wind-shift lines produced dynamically by the pressure-jump line; (iii) rapid linear increase in wind speed on passage of the pressure jump.

  20. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  1. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  2. Gain measurements in CO2 CW low pressure lasers

    International Nuclear Information System (INIS)

    Rodrigues, N.A.S.; Chanes Junior, J.B.; Jayaram, K.

    1983-01-01

    A series of gain measurements in low pressure CO 2 CW laser were performed in order to study the behaviour of a CO 2 laser ampliflier as a function of pressure and discharge current. A theoretical model, based on rate equations is also presented to describe the laser behaviour and the experimental procedure adopted. (C.L.B.) [pt

  3. A system to control low pressure turbine temperatures

    International Nuclear Information System (INIS)

    1980-01-01

    An improved system to control low pressure turbine cycle steam and metal temperatures by governing the heat transfer operation in a moisture separator-reheater is described. The use of the present invention in a pressurized water reactor or a boiling water reactor steam turbine system is demonstrated. (UK)

  4. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

    Kalafati, D.D.; Petrov, S.I.

    1978-01-01

    Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

  5. Upper plenum dump during reflood in PWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Sudo, Yukio; Griffith, Peter.

    1981-01-01

    Upper plenum dump during reflood in a large break loww-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood. The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnetude of water subcooling. (author)

  6. 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.)

  7. 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)

  8. High pressure apparatus for neutron scattering at low temperature

    International Nuclear Information System (INIS)

    Munakata, Koji; Uwatoko, Yoshiya; Aso, Naofumi

    2010-01-01

    Effects of pressure on the physical properties are very important for understanding highly correlated electron systems, in which pressure-induced attractive phenomena such as superconductivity and magnetically ordered non-Fermi liquid have been observed. Up to now, many scientists have developed a lot of high pressure apparatus for each purpose. The characteristic features of various materials and pressure transmitting media for use of high pressure apparatus are reported. Then, two kinds of clamp type high-pressure cell designed for low-temperature neutron diffraction measurements are shown; one is a piston cylinder type high-pressure cell which can be attached to the dilution refrigerator, and the other one is a newly-developed cubic anvil type high-pressure cell which can generate pressure above 7GPa. We also introduce the results of magnetic neutron scattering under pressure on a pressure-induced superconducting ferromagnet UGe 2 in use of the piston cylinder type clamp cell, and those on an iron arsenide superconductor SrFe 2 As 2 in use of the cubic anvil type clamp cell. (author)

  9. Atmospheric-pressure dielectric barrier discharge with capillary injection for gas-phase nanoparticle synthesis

    International Nuclear Information System (INIS)

    Ghosh, Souvik; Liu, Tianqi; Bilici, Mihai; Cole, Jonathan; Huang, I-Min; Sankaran, R Mohan; Staack, David; Mariotti, Davide

    2015-01-01

    We present an atmospheric-pressure dielectric barrier discharge (DBD) reactor for gas-phase nanoparticle synthesis. Nickel nanoparticles are synthesized by homogenous nucleation from nickelocene vapor and characterized online by aerosol mobility measurements. The effects of residence time and precursor concentration on particle growth are studied. We find that narrower distributions of smaller particles are produced by decreasing the precursor concentration, in agreement with vapor nucleation theory, but larger particles and aggregates form at higher gas flow rates where the mean residence time should be reduced, suggesting a cooling effect that leads to enhanced particle nucleation. In comparison, incorporating a capillary gas injector to alter the velocity profile is found to significantly reduce particle size and agglomeration. These results suggest that capillary gas injection is a better approach to decreasing the mean residence time and narrowing the residence time distribution for nanoparticle growth by producing a sharp and narrow velocity profile. (paper)

  10. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  11. 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

  12. A case study of environmental assisted cracking in a low alloy steel under simulated environment of pressurized water reactor

    International Nuclear Information System (INIS)

    Shahzad, M.; Qureshi, A.H.; Waqas, H.; Hussain, N.

    2011-01-01

    Highlights: → We study environmental assisted cracking (EAC) in simulated PWR environment. → The corrosion rate in simulated coolant is low but increases with B conc. → A516 steel shows EAC in simulated coolant particularly at high oxygen levels. → Fracture occurs when the surface cracks join the subsurface cracks. → Corrosion of MnS inclusions and ferrite provide crack nucleation sites. -- Abstract: The electromechanical behavior of a pressure vessel grade steel A516 has been investigated using potentiodynamic polarization curves and slow strain rate test (SSRT) in simulated environment of pressurized water reactor. The anodic polarization behavior shows that the steel remains active in the solution till localized attack (pitting) starts. The cracks initiated at the surface propagate in a trans-granular mode. These cracks are initiated at the inclusion (MnS) sites and at the interfaces between local anode (ferrite) and local cathode (pearlite). It seems that the ultimate fracture occurs when the propagating surface cracks join the subsurface hydrogen induced cracks. The addition of oxygen in the testing chamber to supersaturation levels shifts the corrosion potential to anodic side and significantly lowers the strength and ductility. Compared to the room temperature properties, the UTS and tensile elongation in various simulated conditions are reduced by 10-25% and 25-75%, respectively.

  13. Fuel coolant interaction experiment by direct electrical heating method

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Hirano, Kenmei

    1979-01-01

    In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

  14. Modeling low pressure baroreceptors and their contribution to blood pressure control

    OpenAIRE

    Sánchez de Zambrano, Betsy Mirley; Rojas-Sulbarán, Rubén Darío

    2016-01-01

    The main mechanism for blood pressure (BP) control is coordinated by the central nervous system through the sympathetic and parasympathetic systems. In order to simulate this mechanism, different mathematical models are available, but they take into account only the high pressure receptors as sensing systems for BP. However, other receptors located in low pressure areas have not, as far as we know, been considered in the models described in the literature, despite their important role in the ...

  15. Blood pressure response to low level static contractions

    DEFF Research Database (Denmark)

    Fallentin, Nils; Jørgensen, Kurt

    1992-01-01

    The present study re-examines the 15% MVC concept, i.e. the existence of a circulatory steady-state in low intensity static contractions below 15% of maximal voluntary contraction (MVC). Mean arterial blood pressure was studied during static endurance contractions of the elbow flexor and extensor...... 0.7) min for elbow extension]. Mean arterial blood pressure exhibited a continuous and progressive increase during the 10% MVC contractions indicating that the 15% MVC concept would not appear to be valid. The terminal blood pressure value recorded at the point of exhaustion in the 10% MVC elbow...... the circulation to the muscles was arrested just prior to the cessation of the contraction, blood pressure only partly recovered and remained elevated for as long as the occlusion persisted, indicating the level of pressure-raising muscle chemoreflexes. Based on blood pressure recordings obtained during...

  16. Coolant system decontamination

    International Nuclear Information System (INIS)

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P.

    1981-01-01

    An improved method for decontaminating the coolant system of water cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution. (author)

  17. CT study in primary low spinal fluid pressure syndrome

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Moritoshi; Okayama, Kenji; Kubo, Hiromasa; Watanabe, Hiromi; Endou, Riuko (Ohmiya Red Cross Hospital, Yono, Saitama (Japan))

    1991-02-01

    CT findings in primary low spinal fluid pressure syndrome were studied on the basis of 3 cases. Case 1 was a 43-year-old male with a complicated bilateral isodense subdural hematoma (SDH). Case 2 was a 45-year-old female with a complicated bilateral high dense SDH. Case 3 was a 36-year-old female discharged without any complications after spinal fluid pressure normalized. Slight downward displacement of the brain under low spinal fluid pressure was shown as the narrowing of a Sylvian fissures and infratentorial cisterns on CT. On the other hand, in this syndrome with a complicated bilateral isodense SDH, in addition to this finding, CT revealed distortion and narrowing of body lateral ventricles, which might be differential findings from this syndrome without complicated SDH. Under low spinal fluid pressure, bridging veins are more stretched by a downward displacement of the brain. And consequently they were easily injured and SDH was developed. (author).

  18. The effect of flow direction and magnitude on CHF for low pressure water in thin rectangular channels

    International Nuclear Information System (INIS)

    Mishima, K.; Nishihara, H.

    1985-01-01

    Critical heat flow (CHF) at low flow condition can become important in an MTR-type research reactor under a number of accident conditions. Regardless of the initial stages of these accidents, a condition which is basically the decay heat removal by natural convention boiling can develop. Under such conditions, burnout may occur even at a very low heat flow. In view of this, the CHF at low-flow-rate and low-pressure conditions has been studied for water flowing in thin rectangular channels. Experiments were carried out with two types of rectangular test sections, namely, the one heated from one wide side and the other heated from two opposite sides. In order to observe the effects of gravity, CHF was measured both in upflow and downflow. The CHF at complete bottom blockage was also studied. The results indicate that burnout can occur at a much lower heat flux than pool-boiling CHF or than predicted by the conventional correlations. There was observed a minimum CHF at complete bottom blockage and at very low downflow. The low CHF at very low downflow appears to be due to the stagnation of the bubble in the heated section. This fact indicates that special care should be taken in analyzing the boiling phenomenon which occurs when the coolant flow is very low in a low pressure system. (author)

  19. Verification tests for GRAD, a computer program to predict nonuniform deformation and failure of Zr-2.5 wt percent Nb pressure tubes during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shewfelt, R.S.W.; Godin, D.P.

    1985-03-01

    During a postulated loss-of-coolant accident in a CANDU reactor, the temperature of the pressure tubes could rise sufficiently so that ballooning could occur. It is also likely that there would be a variation in temperature around the tube circumference, causing the deformation to be nonuniform. Since the deformation of the pressure tube controls how the core heat is transferred to the surrounding moderator, which is a large heat sink, a computer program, GRAD, has been developed to predict this nonuniform deformation. Numerous biaxial creep tests were done, where the temperature of internally pressurized sections of Zr-2.5 wt percent Nb pressure tubes were ramped to check the ability of GRAD to predict the resulting nonuniform deformation and possible tube failure. GRAD was successful in predicting the average transverse creep strain observed during the tests and the local transverse creep strain at the end of the tests. GRAD was also able to predict the failure time and average transverse creep strain at failure for all the specimens that failed

  20. Effect of high pressure hydrogen on low-cycle fatigue

    International Nuclear Information System (INIS)

    Rie, K.T.; Kohler, W.

    1979-01-01

    It has been shown that the fatigue life can be influenced in low-cycle range by high pressure hydrogen while the effect of high pressure hydrogen on high-cycle fatigue will not be as significant. The paper reports the details and the results of the investigations of the effect of high pressure hydrogen on the low-cycle endurance of commercially pure titanium. The results of this study indicate that: 1. The degradation of the fatigue life in low-cycle region for commercially pure titanium under high pressure hydrogen can be described by Nsub(cr)sup(α x Δepsilon)sub(pl)sup(=c) 2. The fatigue life decreases with decreasing strain rate. 3. The fatigue life decreases with increasing hydrogen pressure. It was found that the semilogarithmic plot of the fatigue life versus the hydrogen pressure gives a linear relationship. The Sievert's law does not hold in low-cycle fatigue region. 4. HAC in titanium in low-cycle fatigue region is the result of the disolution of hydrogen at the crack tip and of the strain-induced hybride formation. (orig.) 891 RW/orig. 892 RKD [de

  1. Device for preventing coolant in a reactor from being lost

    International Nuclear Information System (INIS)

    Maruyama, Hiromi; Matsumoto, Tomoyuki.

    1975-01-01

    Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

  2. Installation for low temperature vapor explosion experiment

    International Nuclear Information System (INIS)

    Nilsuwankosit, Sunchai; Archakositt, Urith

    2000-01-01

    A preparation for the experiment on the low temperature vapor explosion was planned at the department of Nuclear Technology, Chulalongkorn University, Thailand. The objective of the experiment was to simulate the interaction between the molten fuel and the volatile cooling liquid without resorting to the high temperature. The experiment was expected to involve the injection of the liquid material at a moderate temperature into the liquid material with the very low boiling temperature in order to observe the level of the pressurization as a function of the temperatures and masses of the applied materials. For this purpose, the liquid nitrogen and the water were chosen as the coolant and the injected material for this experiment. Due to the size of the installation and the scale of the interaction, only lumped effect of various parameters on the explosion was expected from the experiment at this initial stage. (author)

  3. Injection therapy for subacute and chronic low back pain: an updated Cochrane review.

    Science.gov (United States)

    Staal, J Bart; de Bie, Rob A; de Vet, Henrica C W; Hildebrandt, Jan; Nelemans, Patty

    2009-01-01

    A systematic review of randomized controlled trials (RCTs). To determine if injection therapy is more effective than placebo or other treatments for patients with subacute or chronic low back pain. The effectiveness of injection therapy for low back pain is still debatable. Heterogeneity of target tissue, pharmacological agent, and dosage, generally found in RCTs, point to the need for clinically valid comparisons in a literature synthesis. We updated the search of the earlier systematic review and searched the Cochrane Central Register of Controlled Trials, MEDLINE, and EMBASE databases up to March 2007 for relevant trials reported in English, French, German, Dutch, and Nordic languages. We also screened references from trials identified. RCTs on the effects of injection therapy involving epidural, facet, or local sites for subacute or chronic low back pain were included. Studies that compared the effects of intradiscal injections, prolotherapy, or ozone therapy with other treatments were excluded unless injection therapy with another pharmaceutical agent (no placebo treatment) was part of one of the treatment arms. Studies about injections in sacroiliac joints and studies evaluating the effects of epidural steroids for radicular pain were also excluded. Eighteen trials (1179 participants) were included in this review. The injection sites varied from epidural sites and facet joints (i.e. intra-articular injections, peri-articular injections and nerve blocks) to local sites (i.e. tender-and trigger points). The drugs that were studied consisted of corticosteroids, local anesthetics, and a variety of other drugs. The methodologic quality of the trials was limited with 10 of 18 trials rated as having a high methodologic quality. Statistical pooling was not possible because of clinical heterogeneity in the trials. Overall, the results indicated that there is no strong evidence for or against the use of any type of injection therapy. There is insufficient evidence to

  4. Use of subconjunctival injections of 5-fluorouracil to rescue and prolong intraocular pressure reduction for a failing Ahmed glaucoma implant.

    Science.gov (United States)

    Kaplowitz, Kevin; Khodadadeh, Sarah; Wang, Samantha; Lee, Daniel; Tsai, James C

    2017-06-01

    5-Fluorouracil (5-FU) has been well described for a failing trabeculectomy bleb, but not for aqueous shunts. We sought to determine whether subconjunctival 5-FU prolongs the intraocular pressure (IOP) efficacy of Ahmed shunts. We included all patients with Ahmed FP-7 implantation by one surgeon at Yale University. Patients with  21 on >2 medications. Five-milligram (0.1 cc) injections were made over the plate. The control group consisted of Ahmed FP-7 patients without injections. The main outcome measure was IOP. Secondary outcome was success (IOP Ahmed shunts. Outcomes between eyes receiving injections and controls were statistically similar.

  5. 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.

  6. Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

    We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

  7. Containment for low temperature district nuclear-heating reactor

    International Nuclear Information System (INIS)

    He Shuyan; Dong Duo

    1992-03-01

    Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

  8. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  9. Paraspinous Lidocaine Injection for Chronic Nonspecific Low Back Pain: A Randomized Controlled Clinical Trial.

    Science.gov (United States)

    Imamura, Marta; Imamura, Satiko Tomikawa; Targino, Rosa Alves; Morales-Quezada, León; Onoda Tomikawa, Luis C; Onoda Tomikawa, Luis G; Alfieri, Fabio M; Filippo, Thais R; da Rocha, Ivan D; Neto, Raul Bolliger; Fregni, Felipe; Battistella, Linamara Rizzo

    2016-05-01

    In this large, sham-controlled, randomized trial, we examined the efficacy of the combination of standard treatment and paraspinous lidocaine injection compared with standard therapy alone in subjects with chronic low back pain. There is little research-based evidence for the routine clinical use of paraspinous lidocaine injection for low back pain. A total of 378 subjects with nonspecific chronic low back pain were randomized to 3 groups: paraspinous lidocaine injection, analgesics, and exercises (group 1, LID-INJ); sham paraspinous lidocaine injection, analgesics, and exercises (group 2, SH-INJ); and analgesics and exercises (group 3, STD-TTR). A blinded rater assessed the study outcomes at 3 time points: baseline, after treatment, and after 3 months of follow-up. There were increased frequency of pain responses and better low back functional scores in the LID-INJ group compared with the SH-INJ and STD-TTR groups. These effects remained at the 3-month follow-up but differed between all 3 groups. There were significant changes in pain threshold immediately after treatment, supporting the effects of this intervention in reducing central sensitization. Paraspinous lidocaine injection therapy is not associated with a higher risk of adverse effects compared with conventional treatment and sham injection. Its effects on hyperalgesia might correlate with changes in central sensitization. NCT02387567. There are few data to support paraspinous lidocaine injection use in patients with nonspecific chronic low back pain. Our results show that this therapy when combined with standard therapy significantly increases the number of responders versus standard treatment alone. Its effects on hyperalgesia might correlate with a change in central sensitization. Copyright © 2016. Published by Elsevier Inc.

  10. Swine manure injection with low-disturbance applicator and cover crops reduce phosphorus losses.

    Science.gov (United States)

    Kovar, J L; Moorman, T B; Singer, J W; Cambardella, C A; Tomer, M D

    2011-01-01

    Injection of liquid swine manure disturbs surface soil so that runoff from treated lands can transport sediment and nutrients to surface waters. We determined the effect of two manure application methods on P fate in a corn (Zea mays L.)-soybean [Glycine max (L.) Merr.] production system, with and without a winter rye (Secale cereale L.)-oat (Avena sativa L.) cover crop. Treatments included: (i) no manure; (ii) knife injection; and (iii) low-disturbance injection, each with and without the cover crop. Simulated rainfall runoff was analyzed for dissolved reactive P (DRP) and total P (TP). Rainfall was applied 8 d after manure application (early November) and again in May after emergence of the corn crop. Manure application increased soil bioavailable P in the 20- to 30-cm layer following knife injection and in the 5- to 20-cm layer following low-disturbance injection. The low-disturbance system caused less damage to the cover crop, so that P uptake was more than threefold greater. Losses of DRP were greater in both fall and spring following low-disturbance injection; however, application method had no effect on TP loads in runoff in either season. The cover crop reduced fall TP losses from plots with manure applied by either method. In spring, DRP losses were significantly higher from plots with the recently killed cover crop, but TP losses were not affected. Low-disturbance injection of swine manure into a standing cover crop can minimize plant damage and P losses in surface runoff while providing optimum P availability to a subsequent agronomic crop.

  11. Enhanced oil recovery by nitrogen and carbon dioxide injection followed by low salinity water flooding for tight carbonate reservoir: experimental approach

    Science.gov (United States)

    Georges Lwisa, Essa; Abdulkhalek, Ashrakat R.

    2018-03-01

    Enhanced Oil Recovery techniques are one of the top priorities of technology development in petroleum industries nowadays due to the increase in demand for oil and gas which cannot be equalized by the primary production or secondary production methods. The main function of EOR process is to displace oil to the production wells by the injection of different fluids to supplement the natural energy present in the reservoir. Moreover, these injecting fluids can also help in the alterations of the properties of the reservoir like lowering the IFTs, wettability alteration, a change in pH value, emulsion formation, clay migration and oil viscosity reduction. The objective of this experiment is to investigate the residual oil recovery by combining the effects of gas injection followed by low salinity water injection for low permeability reservoirs. This is done by a series of flooding tests on selected tight carbonate core samples taken from Zakuum oil field in Abu Dhabi by using firstly low salinity water as the base case and nitrogen & CO2injection followed by low salinity water flooding at reservoir conditions of pressure and temperature. The experimental results revealed that a significant improvement of the oil recovery is achieved by the nitrogen injection followed by the low salinity water flooding with a recovery factor of approximately 24% of the residual oil.

  12. Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (IIV), Part I, IZ-240-o379-1963, Vol. I, Head of the low temperature RA reactor coolant loop

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. Materials for irradiation are metallurgy and chemical samples. In addition to the project objectives, this volume includes technical specifications of the coolant loop head, thermal calculations, calculations of mechanical stress, antireactivity and activation of the construction materials, cost estimation, scheme of the coolant loop head, diagrams of CO 2 gas temperature, thermal neutron flux distribution, design specifications of two proposed solutions for head of low temperature coolant loop [sr

  13. 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.

  14. Low vs Standard Pressures in Gynecologic Laparoscopy: a Systematic Review.

    Science.gov (United States)

    Kyle, Esther B; Maheux-Lacroix, Sarah; Boutin, Amélie; Laberge, Philippe Y; Lemyre, Madeleine

    2016-01-01

    The optimal intraperitoneal pressure during laparoscopy is not known. Recent literature found benefits of using lower pressures, but the safety of doing abdominal surgery with low peritoneal pressures needs to be assessed. This systematic review compares low with standard pneumoperitoneum during gynecologic laparoscopy. We searched Medline, Embase, and the Cochrane Library for randomized controlled trials comparing intraperitoneal pressures during gynecologic laparoscopy. Two authors reviewed references and extracted data from included trials. Risk ratios, mean differences, and standard mean differences were calculated and pooled using RevMan5. Of 2251 studies identified, three were included in the systematic review, for a total of 238 patients. We found a statistically significant but modest diminution in postoperative pain of 0.38 standardized unit based on an original 10-point scale (95% confidence interval [CI], -0.67 to -0.08) during the immediate postoperative period when using low intraperitoneal pressure of 8 mm Hg compared with ≥ 12 mm Hg and of 0.50 (95% CI, -0.80 to -0.21) 24 hours after the surgery. Lower pressures were associated with worse visualization of the surgical field (risk ratio, 10.31; 95% CI, 1.29-82.38). We found no difference between groups over blood loss, duration of surgery, hospital length of stay, or the need for increased pressure. Low intraperitoneal pressures during gynecologic laparoscopy cannot be recommended on the behalf of this review because improvement in pain scores is minimal and visualization of the surgical field is affected. The safety of this intervention as well as cost-effectiveness considerations need to be further studied.

  15. Monitoring of injected CO2 at two commercial geologic storage sites with significant pressure depletion and/or re-pressurization histories: A case study

    Directory of Open Access Journals (Sweden)

    Dayanand Saini

    2017-03-01

    The monitoring technologies that have been used/deployed/tested at both the normally pressured West Hastings and the subnormally pressured Bell Creek storage sites appear to adequately address any of the potential “out of zone migration” of injected CO2 at these sites. It would be interesting to see if any of the collected monitoring data at the West Hastings and the Bell Creek storage sites could also be used in future to better understand the viability of initially subnormally pressured and subsequently depleted and re-pressurized oil fields as secure geologic CO2 storage sites with relatively large storage CO2 capacities compared to the depleted and re-pressurized oil fields that were initially discovered as normally pressured.

  16. Adsorption of helium gas near Tλ at low pressures

    International Nuclear Information System (INIS)

    Kachalin, G.V.; Kryukov, A.P.; Nesterov, S.B.

    1998-01-01

    Cryosorption of helium isotopes ( 4 He and 3 He) on thin argon cryo layers is studied experimentally in the temperature range 4.2-2 K at low pressures. It is shown that the sorption iso stere 4 He is anomalous at temperatures close to be temperature of the phase transition in the bulk of 4 He, T λ . An abrupt pressure change is observed for a 4 He film thickness approximately equal to two monolayers. The experiments on cryosorption of 3 He gas on an argon layer with a 3 He film thickness of approximately one monolayer display monotonous changes in the pressure within the whole temperature range

  17. Evaulation of Incidence and Risk Factors for Intraocular Pressure Elevation After Intravitreal Triamcinolone Acetonide Injection

    Directory of Open Access Journals (Sweden)

    Didar Uçar

    2015-05-01

    Full Text Available Objectives: To investigate the effect of intravitreal triamcinolone acetonide (IVTA used for the macular edema on intraocular pressure (IOP and to determine the risk factors for IOP elevation. Materials and Methods: This retrospective study included 93 eyes of 85 patients who had 4 mg intravitreal triamcinolone injection. Of the 85 patients, 56 (65.8% had diabetic macular edema, 22 (25.8% had branch retinal, and 7 (8.2% had central retinal vein occlusion. IOP changes after injection as well as the relation between IOP elevation and age, sex, lens status, etiology of macular edema, baseline IOP were evaluated. Results: Fourty-six male and 39 female patients with mean age 61.58±9.5 years were evaluated. IOP was recorded to be >24 mmHg in 30 eyes (32.2% at follow-up visit after an average of 7.5 weeks. Normalization of IOP with medication was achieved in all IOP elevated eyes. Fifteen of 29 eyes (51.7% with vein occlusion and 15 of 64 eyes (23.3% with diabetic macula edema had IOP elevation (p=0.01. Twenty-six of 73 phakic (35.6% and 4 of 20 pseudophakic eyes (20% had IOP >24 mmHg (p=0.16. There was no association between IOP elevation and sex (p=0.33. Baseline IOP was 16.47±2.8 mmHg in eyes which had elevated IOP and 14.78±2.4 mmHg in the remaining. There was significant relation between IOP elevation and baseline IOP level (p=0.01. Conclusion: Elevated IOP is common side effect after IVTA, but normalization is usually achieved by topical medication. Patients with baseline IOP ≥15 mmHg and vein occlusion have higher risk for IOP elevation. (Turk J Ophthalmol 2015; 45: 86-91

  18. Reactor coolant pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Harand, E.; Richter, G.; Tschoepel, G.

    1975-01-01

    A brake for the pump rotor of a main coolant pump or a shutoff member on the pump are provided in order to prevent excess speeds of the pump rotor. Such excess speeds may occur in PWR type reactors with water at a pressure below, e.g., 150 bars if there is leakage from a coolant line associated with the main coolant pump. As a brake, a centrifugal brake depending upon the pump speed or a brake ring arranged on the pump housing and acting on the pump rotor, which ring would be activated by pressure differentials in the pump, may be used. If the pressure differences between suction and pressure sockets are very small, a controlled hydraulic increase of the pressure force on the brake may also be provided. Furthermore, a turbine brake may be provided. A slide which is automatically movable in closing position along the pump rotor axis is used as a shutoff element. It is of cylindrical configuration and is arranged concentrically with the rotor axis. (DG) [de

  19. Hydrodynamics of Safety Injection Tank with Fluidic Device in Recent Regulation

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Yoo, Seung Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Safety Injection Tank (SIT) with Fluidic Device (FD) has been used in several APR1400 nuclear power plants. It was designed to provide a longer passive safety injection than the existing accumulator to improve the safety for Large Break Loss-of-Coolant Accident (LBLOCA) by changing the injected flow through the FD and the standpipe of the SIT. As a result, high flow injection phase and the subsequent low flow one can be achieved as longer than the existing accumulator. The present paper discusses the major concerns related to SIT hydrodynamics and the directions to resolution recently concerned. Modeling of SIT/FD by total hydraulic resistances, potential of nitrogen intrusion, and effect of initial pressure of SIT testing are included. Based on the discussion, a table of the important phenomena of the SIT/FD was proposed with the relevancy of the calculation models applied. The present paper discussed the SIT hydrodynamics including the modeling of SIT/FD by total hydraulic resistances, potential of nitrogen intrusion, and effect of initial pressure of SIT testing. Also a table of the important phenomena of the SIT/FD was proposed with the relevancy of the calculation models applied. The following conclusions are obtained uncertainty due to the assumption of the total Kfactor as constant for high flow, transition phase, and low flow phase should be considered and nitrogen intrusion phenomena during the transition phase should be considered with a conservatism, especially considering the current situation of nonmeasuring the standpipe level.

  20. Hydrodynamics of Safety Injection Tank with Fluidic Device in Recent Regulation

    International Nuclear Information System (INIS)

    Bang, Young Seok; Yoo, Seung Hun

    2016-01-01

    Safety Injection Tank (SIT) with Fluidic Device (FD) has been used in several APR1400 nuclear power plants. It was designed to provide a longer passive safety injection than the existing accumulator to improve the safety for Large Break Loss-of-Coolant Accident (LBLOCA) by changing the injected flow through the FD and the standpipe of the SIT. As a result, high flow injection phase and the subsequent low flow one can be achieved as longer than the existing accumulator. The present paper discusses the major concerns related to SIT hydrodynamics and the directions to resolution recently concerned. Modeling of SIT/FD by total hydraulic resistances, potential of nitrogen intrusion, and effect of initial pressure of SIT testing are included. Based on the discussion, a table of the important phenomena of the SIT/FD was proposed with the relevancy of the calculation models applied. The present paper discussed the SIT hydrodynamics including the modeling of SIT/FD by total hydraulic resistances, potential of nitrogen intrusion, and effect of initial pressure of SIT testing. Also a table of the important phenomena of the SIT/FD was proposed with the relevancy of the calculation models applied. The following conclusions are obtained uncertainty due to the assumption of the total Kfactor as constant for high flow, transition phase, and low flow phase should be considered and nitrogen intrusion phenomena during the transition phase should be considered with a conservatism, especially considering the current situation of nonmeasuring the standpipe level

  1. CFD simulation of subcooled flow boiling at low pressure

    International Nuclear Information System (INIS)

    Koncar, B.; Mavko, B.

    2001-01-01

    An increased interest to numerically simulate the subcooled flow boiling at low pressures (1 to 10 bar) has been aroused in recent years, pursued by the need to perform safety analyses of research nuclear reactors and to investigate the sump cooling concept for future light water reactors. In this paper the subcooled flow boiling has been simulated with a multidimensional two-fluid model used in a CFX-4.3 computational fluid dynamics (CFD) code. The existing model was adequately modified for low pressure conditions. It was shown that interfacial forces, which are usually used for adiabatic flows, need to be modeled to simulate subcooled boiling at low pressure conditions. Simulation results are compared against published experimental data [1] and agree well with experiments.(author)

  2. Epidural blood patch for refractory low CSF pressure headache

    DEFF Research Database (Denmark)

    Madsen, Søren Aalbæk; Fomsgaard, Jonna Storm; Jensen, Rigmor

    2011-01-01

    primary effect parameter was total headache burden defined as area under the curve (AUC: intensity × duration) and as secondary effect parameters we identified: intensity (VAS 0-10), frequency (days per month), duration in hours (total hours/month) and also medication days (days on medication...... of non-invasive/conservative measures and invasive measures with epidural blood patch providing the cornerstone of the invasive measures. In the present pilot study we therefore aimed to evaluate the treatment efficacy of epidural blood patch (EBP) in treatment-refractory low-pressure headache. Our......Once believed an exceedingly rare disorder, recent evidence suggests that low cerebrospinal fluid (CSF) pressure headache has to be considered an important cause of new daily persistent headaches, particularly among young and middle-aged individuals. Treatment of low CSF pressure headache consists...

  3. Epidural blood patch for refractory low CSF pressure headache

    DEFF Research Database (Denmark)

    Madsen, Søren Aalbæk; Fomsgaard, Jonna Storm; Jensen, Rigmor

    2011-01-01

    of non-invasive/conservative measures and invasive measures with epidural blood patch providing the cornerstone of the invasive measures. In the present pilot study we therefore aimed to evaluate the treatment efficacy of epidural blood patch (EBP) in treatment-refractory low-pressure headache. Our......Once believed an exceedingly rare disorder, recent evidence suggests that low cerebrospinal fluid (CSF) pressure headache has to be considered an important cause of new daily persistent headaches, particularly among young and middle-aged individuals. Treatment of low CSF pressure headache consists...... primary effect parameter was total headache burden defined as area under the curve (AUC: intensity × duration) and as secondary effect parameters we identified: intensity (VAS 0-10), frequency (days per month), duration in hours (total hours/month) and also medication days (days on medication...

  4. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

    Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

    1991-01-01

    In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

  5. 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.

  6. Ideal magnetohydrodynamic simulations of low beta compact toroid injection into a hot strongly magnetized plasma

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Wei [Los Alamos National Laboratory; Hsu, Scott [Los Alamos National Laboratory; Li, Hui [Los Alamos National Laboratory

    2009-01-01

    We present results from three-dimensional ideal magnetohydrodynamic simulations of low {beta} compact toroid (CT) injection into a hot strongly magnetized plasma, with the aim of providing insight into CT fueling of a tokamak with parameters relevant for ITER (International Thermonuclear Experimental Reactor). A regime is identified in terms of CT injection speed and CT-to-background magnetic field ratio that appears promising for precise core fueling. Shock-dominated regimes, which are probably unfavorable for tokamak fueling, are also identified. The CT penetration depth is proportional to the CT injection speed and density. The entire CT evolution can be divided into three stages: (1) initial penetration, (2) compression in the direction of propagation and reconnection, and (3) coming to rest and spreading in the direction perpendicular to injection. Tilting of the CT is not observed due to the fast transit time of the CT across the background plasma.

  7. Testing of low pressure proton exchange membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Bettoni, M; Naso, V; Lucentini, M; Rubini, L

    1998-07-01

    One of the main issues concerning PEMFC is the choice of operating pressure, for both stationary and automotive applications. This is because the air compressor may absorb a significant amount--up to 25%--of the power output of the fuel cells stack. A comparison has been made between the performance of various stacks of different dimensions, tested in the De Nora Laboratories operated at high (4 bar) and low (1.5 bar) pressures, considering power output reduced by the compressor power absorption. Differences of performance and efficiency between high and low pressure stacks have been noticed in the range of 10%. In operating at low pressure, higher efficiency is obtainable, but the maximum power of the stack is less; this means less fuel consumption, but requires a greater reacting surface and larger dimension of the stack. Consequently low pressures make the system simpler (a blower can be used instead of a compressor), and safer (there is practically no risk of breaking the membrane).

  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. Minimum-Voltage Vector Injection Method for Sensorless Control of PMSM for Low-Speed Operations

    DEFF Research Database (Denmark)

    Xie, Ge; Lu, Kaiyuan; Kumar, Dwivedi Sanjeet

    2016-01-01

    In this paper, a simple signal injection method is proposed for sensorless control of PMSM at low speed, which ideally requires one voltage vector only for position estimation. The proposed method is easy to implement resulting in low computation burden. No filters are needed for extracting...... may also be further developed to inject two opposite voltage vectors to reduce the effects of inverter voltage error on the position estimation accuracy. The effectiveness of the proposed method is demonstrated by comparing with other sensorless control method. Theoretical analysis and experimental...

  10. Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III

    International Nuclear Information System (INIS)

    Koizumi, Yasuo; Tasaka, Kanji

    1986-01-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25 % main recirculation pump suction line break (MRPS-B) experiments, the 21 % single-ended jet pump drive line break (JPD-B) experiment and the 15 % main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop. (author)

  11. Implementation of pressurized air injection system in a Kaplan prototype for the reduction of vibration caused by tip vortex cavitation

    Science.gov (United States)

    Rivetti, A.; Angulo, M.; Lucino, C.; Hene, M.; Capezio, O.; Liscia, S.

    2016-11-01

    Blade tip cavitation is a well-known phenomenon that affects the performance of large-diameter Kaplan turbines and induces structural vibration. Injection of pressurized air has been found to yield promising results in reducing those damaging effects. In this work, the results of an experimental test of air injection on a 9.5-m-diameter Kaplan turbine are reported. Experiments were performed for several load conditions and for two different net heads. Accelerations, pressure pulsation and noise emission were monitored for every tested condition. Results show that, at the expense of a maximum efficiency drop of 0.2%, air injection induces a decrease on the level of vibration from 57% up to 84%, depending on the load condition. Such decrease is seen to be proportional to the air flow rate, in the range from 0.06 to 0.8‰ (respect to the discharge at the best efficiency point).

  12. Effect of pegaptanib sodium 0.3 mg intravitreal injections (Macugen) in intraocular pressure: posthoc analysis from VISION study

    NARCIS (Netherlands)

    Boyer, David S.; Goldbaum, Mauro; Leys, Anita M.; Starita, Carla; Blumenkranz, M.; Buyse, M.; Goldberg, M.; Gragoudas, E. S.; Miller, J.; Schwartz, S. D.; Singerman, L.; Yannuzzi, L.; Adamis, A. P.; Guyer, D. R.; O'Shaughnessy, D.; de Gronckel, S.; Fesneau, G.; Quinaux, E.; Tremolet, D.; Wang, K.; Brailey, A.; Finman, J.; Ting, N.; Bressler, N. M.; Bressler, S. B.; Denblow, R.; Schein, O. D.; Seabrook, S.; Solomon, S.; Schachat, A. P.; Philips, D.; Altaweel, M.; Davis, M. D.; Blodi, Ba; Danis, R. P.; Ip, M. S.; Hiner, C.; Elledge, J.; Webster, M.; Hannan, C.; Ficken, J.; Alexander, S.; Neider, M.; Wabers, H.; Vargo, P.; Lambert, E.; Kastorff, L.; Carr, A.; Shkiele, A.; Schlingemann, R. O.

    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 the

  13. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Patel, R J [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

    1994-12-31

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs.

  14. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    International Nuclear Information System (INIS)

    Mukherjee, S.; Ghosh, J.K.; Patel, R.J.

    1994-01-01

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs

  15. Durability of low-pH injection grout. A literature survey

    International Nuclear Information System (INIS)

    Holt, E.

    2008-01-01

    This publication provides an overview of the durability of injection grouts. It is intended for use during planning and construction at the ONKALO underground research facility. The review has been done with respect to the application conditions, materials and service life requirements expressed by Posiva Oy. The publication describes all types of cement-based material durability, with an emphasis on the key issues of shrinkage cracking, leaching and sulphate attack. The second part of the report provides information on how durability expectations have changed with the history of injection grout development. The report gives information specific to low-pH injection grouts containing high amounts of silica fume performance and how their durability is expected to differ from traditional normal cement-based mixtures. The final part of the report provides suggestions for future research needs for ensuring the service life of injection grouts. The key finding from this study is that the low-pH grout material is not expected to have worse durability performance compared to the standard injection grout. Combining high amounts of silica fume with the cement to produce low-pH grout should result in a material having lower permeability and thus greater resistance to leaching and chemical attack. Further laboratory testing is needed to quantitatively verify these findings and to provide inputs for future service life modeling. (orig.)

  16. FY 2000 report on the results of the technology development of energy use reduction of machine tools, etc. Technology development of environmental load reduction related to water soluble lubricating oil, etc. (R and D of low energy coolant degradation prevention technology and waste liquid processing technology); 2000 nendo energy shiyo gorika kosaku kikai nado gijutsu kaihatsu seika hokokusho. Suiyosei junkatsuyu ni kakawaru kankyo fuka teigen nado gijutsu kanri (tei energy coolant fuhai boshi gijutsu oyobi haieki shori gijutsu no kenkyu kaihatsu)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    The R and D were carried out on a system by which the long life of coolant of machine tools can be achieved and a system to process waste efficiently, economically and with less environmental loads, and the FY 2000 results were summed up. In the R and D of a system to prevent degradation of low energy coolant, measurement was made of effects of