RIA Analysis of Unprotected TRIGA Reactor

Directory of Open Access Journals (Sweden)

M.H. Altaf

2017-07-01

Full Text Available An RIA (reactivity initiated accident analysis has been carried out for the TRIGA Mark II research reactor considering both step and ramp reactivity ranges within 0.5 % dk/k (< $1 to 2.0 % dk/k (>$2. The insertion time was set at 10 s. Based on the fact that a reactor becomes unprotected if scram does not work at the event of danger, to define unprotected conditions, the time to actuate scram (trip was taken as close to total simulation time. In this long duration of scram inactivity, it is obtained from the present analysis that the reactor remained safe to up to 1.8 % dk/k ($2.57 for step reactivity and 1.99 % dk/k ($2.84 for ramp reactivity. In addition to negative temperature coefficient of reativity, probably the longer time of reactivity insertion keeps TRIGA safe even at larger magnitudes of reactivity during unprotected reactor transients. Coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/R has been utilized for this work. It appears that EUREKA-2/RR predicts the sequence of unprotected transient scenario of TRIGA core with good approximation and the results will definitely be helpful for the reactor operators.

Comparative analysis of unprotected loss-of-flow accidents for the 1.0 m EFR-LVC core using different computer codes

International Nuclear Information System (INIS)

Royl, P.; Frizonnet, J.M.; Moran, J.

1993-02-01

A comparative analysis of the unprotected loss of flow (ULOF) accident has been performed for the LVC core (Lower Void Core) of the European Fast Reactor EFR with the FRAX5B and FRAX5C codes from the AEA-T, the PHYSURAC code from CEA and the SAS4A REF92 code system developed jointly between KfK, CEA and PNC. The accident is triggered by the run down of the coolant pumps with failure to trip the reactor by the primary and/or secondary shutdown system. Only a limited amount of mitigating reactivity from the third shutdown line was considered so that the accident can progress into boiling and core disruption. This code outlines the important modelling differences and compares the different simulations. The discussion of the rather wide spectrum of calculated accident progressions identifies the generic differences, relates them to the applied models, and summarizes the key points that are responsible for the different progressions. A comparison of the consequence spectrum from all simulations indicates zero work energies for the majority of the calculations. All simulations show up the need for a continued accident analysis into the early and late transition phase

Current capabilities of transient two-phase flow instruments

International Nuclear Information System (INIS)

Solbrig, C.W.; Kondic, N.N.

1979-01-01

The measurement of two phase flow phenomena in transient conditions representative of a Loss-of-Coolant Accident requires the use of sophisticated instruments and the further development of other instruments. Measurements made in large size pipes are often flow regime dependent. The flow regimes encountered depend upon the system geometry, transient effects, heat transfer, etc. The geometries in which these measurements must be made, the instruments which are currently used, new instruments being developed, the facilities used to calibrate these instruments, and the improvements which must be made to measurement capabilities are described

Transient analyses for a molten salt fast reactor with optimized core geometry

Energy Technology Data Exchange (ETDEWEB)

Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

2015-10-15

Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

Transient thermal hydraulic analysis of the IAEA 10 MW MTR reactor during Loss of Flow Accident to investigate the flow inversion

International Nuclear Information System (INIS)

AL-Yahia, Omar S.; Albati, Mohammad A.; Park, Jonghark; Chae, Heetaek; Jo, Daeseong

2013-01-01

Highlights: • Transient analyses of a slow and fast LOFA were investigated. • A reactor kinetic and thermal hydraulic coupled model was developed. • Based on force balance, the flow rate during flow inversion was determined. • Flow inversion in a hot channel occurred earlier than in an average channel. • Two temperature peaks were observed during both slow and fast LOFA. - Abstract: Transient analyses of the IAEA 10 MW MTR reactor are investigated during a fast and slow Loss of Flow Accident (LOFA) with a neutron kinetic and thermal hydraulic coupling model. A spatial-dependent thermal hydraulic technique is adopted for analyzing the local thermal hydraulic parameters and hotspot location during a flow inversion. The flow rate through the channel is determined in terms of a balance between driving and preventing forces. Friction and buoyancy forces act as resistance of the flow before a flow inversion while buoyancy force becomes the driving force after a flow inversion. By taking into account the buoyancy effect to determine the flow rate, the difference in the flow inversion time between hot and average channels is investigated: a flow inversion occurs earlier in the hot channel than in an average channel. Furthermore, the movement of the hotspot location before and after a flow inversion is investigated for a slow and fast LOFA. During a flow inversion, two temperature peaks are observed: (1) the first temperature peak is at the initiation of the LOFA, and (2) the second temperature peak is when a flow inversion occurs. The maximum temperature of the cladding is found at the second temperature peak for both LOFA analyses, and is lower than the saturation temperature

A calculation method for transient flow distribution of SCWR(CSR1000)

International Nuclear Information System (INIS)

Chen, Juan; Zhou, Tao; Chen, Jie; Liu, Liang; Muhammad, Ali Shahzad; Muhammad, Zeeshan Ali; Xia, Bangyang

2017-01-01

The supercritical water reactor CSR1000 is selected for the study. A parallel channel flow transient flow distribution module is developed, which is used for solving unsteady nonlinear equations. The incorporated programs of SCAC-CSR1000 are executed on normal and abnormal operating conditions. The analysis shows that: 1. Transient flow distribution can incorporate parallel channel flow calculation, with an error less than 0.1%; 2. After a total loss of coolant flow, the flow of each channel shows a downward trend; 3. In the event of introducing a traffic accident, the first coolant flow shows an increasing trend.

Transient analyses for accelerator driven system PDS-XADS using the extended SIMMER-III code

International Nuclear Information System (INIS)

Suzuki, Tohru; Chen, Xue-Nong; Rineiski, Andrei; Maschek, Werner

2005-01-01

Transient analyses for Preliminary Design Studies of an Experimental Accelerator Driven System (PDS-XADS) were performed with the reactor safety analysis code SIMMER-III, which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors so as to describe the XADS specifics such as subcritical core, strong external neutron source and lead-bismuth-eutectic (LBE) coolant. As transient scenarios, the following cases were analyzed in accordance with the PDS-XADS program: spurious beam trip (BT), unprotected beam overpower (UBOP), unprotected transient overpower (UTOP), unprotected loss of flow (ULOF) and unprotected blockage (UBL) in a single fuel assembly. In addition, to cover some core-melt situations and investigate the potential for recriticalities, so-called snap-shot analyses with ad hoc postulated severe blockage conditions were also investigated. The simulation results for BT and UBOP showed that immediate fuel damage might not take place under short-time beam interruption or a 100% increase of the external neutron source. Concerning UTOP, it was found that a reactivity jump of 1 $ would not lead to damage of the fuel and the cladding. The ULOF simulation showed that the remaining natural convection of the coolant would prevent the cladding from disruptions. In the simulation of UBL in a single fuel assembly, it was shown that no cladding failure might be expected, due to the radial heat transfer and the coolant flow in the hexcan gap. Under an artificial suppression of the radial heat transfer for this UBL case, a pin failure occurred in the simulation but subsequent fuel sweep-out into the upper plenum region would bring a reactivity reduction and no power excursion. The severe accident simulations starting from postulated blockage above an already disrupted core showed that a severe recriticality could be avoided by the fuel sweep-out into the dummy-assembly or hexcan gap regions. The present

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

International Nuclear Information System (INIS)

Shin, Andong; Choi, Yong Won

2016-01-01

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Pitot tube and drag body measurements in transient steam--water flows

International Nuclear Information System (INIS)

Fincke, J.R.; Deason, V.A.; Dacus, M.W.

1979-01-01

The use of full-flow drag devices and rakes of water-cooled Pitot tubes to measure the transient two-phase mass flow during loss-of-coolant experiments in pressurized water reactor (PWR) environments has been developed. Mass flow rate measurements have been obtained in high temperature and pressure environments, similar to PWRs, under transient conditions. Comparisons of the measured time integrated value of mass flow to the known system mass before depressurization are made

Transient and steady-state flows in shock tunnels

Energy Technology Data Exchange (ETDEWEB)

Hannemann, K. [Deutsche Forschungsanstalt fuer Luft- und Raumfahrt e.V. (DLR), Goettingen (Germany); Jacobs, P.A. [Queensland Univ., Brisbane (Australia). Dept. of Mechanical Engineering; Thomas, A.; McIntyre, T.J. [Queensland Univ., Brisbane, QLD. (Australia). Dept. of Physics

1999-12-01

Due to the difficulty of measuring all necessary flow quantities in the nozzle reservoir and the test section of high enthalpy shock tunnels, indirect computational methods are necessary to estimate the required flow parameters. In addition to steady state flow computations of the nozzle flow and the flow past wind tunnel models it is necessary to investigate the transient flow in the facility in order to achieve a better understanding of its performance. These transient effects include the nozzle starting flow, the interaction of the shock tube boundary layers and the reflected shock, thermal losses in the shock reflection region and the developing boundary layers in the expanding section of the nozzle. Additionally, the nonequilibrium chemical and thermal relaxation models which are used to compute high enthalpy flows have to be validated with appropriate experimental data. (orig.)

Transient and steady-state flows in shock tunnels

Energy Technology Data Exchange (ETDEWEB)

Hannemann, K. (Deutsche Forschungsanstalt fuer Luft- und Raumfahrt e.V. (DLR), Goettingen (Germany)); Jacobs, P.A. (Queensland Univ., Brisbane (Australia). Dept. of Mechanical Engineering); Thomas, A.; McIntyre, T.J. (Queensland Univ., Brisbane, QLD. (Australia). Dept. of Physics)

1999-01-01

Due to the difficulty of measuring all necessary flow quantities in the nozzle reservoir and the test section of high enthalpy shock tunnels, indirect computational methods are necessary to estimate the required flow parameters. In addition to steady state flow computations of the nozzle flow and the flow past wind tunnel models it is necessary to investigate the transient flow in the facility in order to achieve a better understanding of its performance. These transient effects include the nozzle starting flow, the interaction of the shock tube boundary layers and the reflected shock, thermal losses in the shock reflection region and the developing boundary layers in the expanding section of the nozzle. Additionally, the nonequilibrium chemical and thermal relaxation models which are used to compute high enthalpy flows have to be validated with appropriate experimental data. (orig.)

Transient flow combustion

Science.gov (United States)

Tacina, R. R.

1984-01-01

Non-steady combustion problems can result from engine sources such as accelerations, decelerations, nozzle adjustments, augmentor ignition, and air perturbations into and out of the compressor. Also non-steady combustion can be generated internally from combustion instability or self-induced oscillations. A premixed-prevaporized combustor would be particularly sensitive to flow transients because of its susceptability to flashback-autoignition and blowout. An experimental program, the Transient Flow Combustion Study is in progress to study the effects of air and fuel flow transients on a premixed-prevaporized combustor. Preliminary tests performed at an inlet air temperature of 600 K, a reference velocity of 30 m/s, and a pressure of 700 kPa. The airflow was reduced to 1/3 of its original value in a 40 ms ramp before flashback occurred. Ramping the airflow up has shown that blowout is more sensitive than flashback to flow transients. Blowout occurred with a 25 percent increase in airflow (at a constant fuel-air ratio) in a 20 ms ramp. Combustion resonance was found at some conditions and may be important in determining the effects of flow transients.

State of the art of CATHARE model for transient safety analysis of ASTRID SFR

International Nuclear Information System (INIS)

Lavastre, R.; Conti, A.; Marsault, Ph.; Chenaud, M.S.; Tosello, A.

2014-01-01

Within the framework of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), the conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves enhancing the general design in order to : - increase the safety margins for all unprotected-loss-of-flow (ULOF) and unprotected-loss-of-heat-sink (ULOHS) transients, - identify the need for additional safety devices that would complement core natural behavior so that temperature criteria on coolant, core and primary circuit structures can remain under the safety criteria. For this purpose, the use of CATHARE system code has been very important from the early stage of design in order to ensure a feedback for design teams to improve behavior during unprotected transients. Until 2012, CATHARE ULOxx transient calculations have been used mainly to compare different core designs. They contributed to lead to the choice of CFV core (axially heterogeneous core with an upper sodium plenum employed to achieve a negative sodium void reactivity worth). Meanwhile, models for an accurate core description and transients have been developed in CATHARE to improve the calculations towards best estimate calculations for safety analysis. This paper therefore presents these main developments in core modeling achieved for the 2 past years. For instance, we will focus on the way of dealing with fuel assemblies that have to be grouped together in the CATHARE code to form a channel with similar neutronic physics and thermal-hydraulics characteristics. We will also explain the way we deal with heterogeneity of fuel pin to obtain the accurate fuel temperature along the axis and to take into account pellet-cladding gap state. These two points have a great importance on feedback effects linked to the fuel, mainly the Doppler effect. The paper will finally introduce the upcoming improvements that are under development nowadays

Gulping phenomena in transient countercurrent two-phase flow

International Nuclear Information System (INIS)

Tehrani, Ali A.K.

2001-04-01

Apart from previous work on countercurrent gas-liquid flow, transient tank drainage through horizontal off-take pipes is described, including experimental procedure, flow pattern on observations and countercurrent flow limitation results. A separate chapter is devoted to countercurrent two-phase flow in a pressurised water reactor hot-leg scaled model. Results concerning low head flooding, high head and loss of bowl flooding, transient draining of the steam generator and pressure variation and bubble detachment are presented. The following subjects are covered as well: draining of sealed tanks of vertical pipes, unsteady draining of closed vessel via vertical tube, unsteady filling of a closed vessel via vertical tube from a constant head reservoir. Practical significance of the results obtained is discussed

Cladding motion and entrainment during loss of flow in the LMFBR

International Nuclear Information System (INIS)

Scale, T.J.; Eggen, D.T.

1978-01-01

The study that has been undertaken here is to characterize the loss of flow with appropriate dimensionless numbers such as Re, sup(p)vap/Psteel, sup(Dh/L), and to suggest a minimum Re, necessary for entrainment of simulant cladding material under transient melting. In particular, we are interested in the downward flow of vapor over an intact rod undergoing a loss of flow or loss of heat sink. (author)

Transient two-phase flow

International Nuclear Information System (INIS)

Hsu, Y.Y.

1974-01-01

The following papers related to two-phase flow are summarized: current assumptions made in two-phase flow modeling; two-phase unsteady blowdown from pipes, flow pattern in Laval nozzle and two-phase flow dynamics; dependence of radial heat and momentum diffusion; transient behavior of the liquid film around the expanding gas slug in a vertical tube; flooding phenomena in BWR fuel bundles; and transient effects in bubble two-phase flow. (U.S.)

Germany: Assessment of the efficiency of a passive safety system for prevention of severe accidents for SFR

International Nuclear Information System (INIS)

Bubelis, E.

2015-01-01

The aim of the study was the evaluation of severe transient behavior in Sodium-cooled Fast Reactor (SFR) and of the impact of newly conceived inherent mitigation measures (the use of ASD – additional shutdown device). The SFR design taken for the analysis was the SFR(v2b-ST) reactor design, and the system code to be used was selected to be the SIM-SFR code. The transients chosen for evaluation of the efficiency of mitigation measures were the unprotected loss-of-flow (ULOF) and the unprotected loss-of-heat-sink (ULOHS)

Effect of automatic recirculation flow control on the transient response for Lungmen ABWR plant

Energy Technology Data Exchange (ETDEWEB)

Tzang, Y.-C., E-mail: yctzang@aec.gov.t [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China); Chiang, R.-F.; Ferng, Y.-M.; Pei, B.-S. [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China)

2009-12-15

In this study the automatic mode of the recirculation flow control system (RFCS) for the Lungmen ABWR plant has been modeled and incorporated into the basic RETRAN-02 system model. The integrated system model is then used to perform the analyses for the two transients in which the automatic RFCS is involved. The two transients selected are: (1) one reactor internal pump (RIP) trip, and (2) loss of feedwater heating. In general, the integrated system model can predict well the response of key system parameters, including neutron flux, steam dome pressure, heat flux, RIP flow, core inlet flow, feedwater flow, steam flow, and reactor water level. The transients are also analyzed for manual RFCS case, between the automatic RFCS and the manual RFCS cases, comparisons of the transient response for the key system parameter show that the difference of transient response can be clearly identified. Also, the results show that the DELTACPR (delta critical power ratio) for the transients analyzed may not be less limiting for the automatic RFCS case under certain combination of control system settings.

Trace analysis of loss of feedwater flow event in Lungmen ABWR

International Nuclear Information System (INIS)

Wang Jongrong; Lin Haotzu; Wang Weichen; Yang Shuming; Shih Chunkuan

2009-01-01

TRACE (TRAC/RELAP Advanced Computational Engine) model of Lungmen Nuclear Power Plant was used to analyze the Loss of Feedwater Flow transient as defined in Lungmen FSAR Chapter 15. The results were compared with those from FSAR and RETRAN02. Lungmen TRACE model will have two models: In model A, vessel is divided into 11 axial levels, 4 radial rings and 1 azimuthal sectors; In model B, vessel is divided into 11 axial levels, 4 radial rings, and 6 azimuthal sectors. The above models include feedwater control system, narrow range water level control system, and wide range water level control system. The loss of feedwater flow (LOFW) transient began with the trip of two operating feedwater pumps either from the pump mechanical/electric failure, or the operator human error, or high water level signal. Feedwater flow was assumed to descend to 0 in 5 seconds and led to the decrease of reactor water level. At L3 low water level setpoint, the system actuated reactor scram signal and RIP trip signal for RIPs not connected to the M/G set. At L2 low-low water level setpoint, the system would trip the other six RIPs. This paper compares those important thermal parameters at steady state, such as the dome pressure and temperature of reactor vessel, steam flow, feedwater flow, core flow, and RIP flow, etc.. It also compares system parameters under transient conditions, such as core thermal power, core flow, steam flow, feedwater flow, Narrow Range Water Level (NRWL), Wide Range Water Level (WRWL) and RIP flow, etc.. It was concluded that the steady state and transient results of TRACE calculations are in good agreement with those from RETRAN02. In summary, our studies concluded that Lungmen TRACE model is correct and accurate enough for future safety analysis applications. (author)

Analysis of unscrammed loss of flow and heat sink for PRISM with GEM

International Nuclear Information System (INIS)

Slovik, G.C.; Van Tuyle, G.J.; Kennett, R.J.

1991-01-01

The US Department of Energy is sponsoring an advanced liquid-metal reactor design by General Electric Company (GE) called PRISM. The intent is to design a reactor with passively safe responses to many operational and severe accidents. PRISM is under review at the US Nuclear Regulatory Commission for licensability with Brookhaven National Laboratory providing technical assistance. Recently, the PRISM design has been modified to include three gas expansion modules (GEMs) on its core periphery. The GEMs were added to quickly reduce the power (by inserting negative reactivity) during loss-of-flow events to curtail peak fuel and clad temperatures predicted in the previous design. The GEM prototypes have been tested at the Fast Flux Test Facility. The significance of the GEMs in PRISM is discussed in this paper through the evaluation of the unprotected loss of flow (ULOF) and loss of heat sink using the SSC code. It has been demonstrated in the past that SSC predicts results similar to GE and other liquid-metal reactor codes

PBDOWN: A computer code for simulation of core material discharge and expansion in the upper coolant plenum in a hypothetical unprotected loss of flow accident in a LMFBR

International Nuclear Information System (INIS)

Royl, P.

1985-01-01

The report gives a description of the code PBDOWN (Pool Blow Down), its equations, input specifications and subroutines and it lists the input and output for some samples. Besides that some analysis results for the SNR-300 are discussed, that were obtained with this code. PBDOWN is an integral blow-down and expansion code, which simulates core material discharge and expansion into a sodium filled upper coolant plenum after build-up of vapour pressures in an unprotected loss of flow accident. The model includes the effect of sodium entrainment into an expending bubble of fuel or steel vapour with various assumptions for the heat transfer and vaporization of the entrained sodium droplets. The expanding vapour bubble is connected to the discharging pool via an orifice of a given size through which a time dependent ejection is simulated using quasi-stationary blow down correlations. The model allows bounding analysis of the possible influence of sodium vapour as a secondary working fluid, that is activated outside the pool on the overall expansion energy and discharge

Control-rod, pressure and flow-induced accident and transient analysis of a direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor

International Nuclear Information System (INIS)

Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki

1996-01-01

The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed

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

International Nuclear Information System (INIS)

Cahalan, J.E.

1986-01-01

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

Flow transients experiments with refrigerant-12

International Nuclear Information System (INIS)

Celata, G.P.; D'Annibale, F.; Farello, G.E.; Setaro, T.

1986-01-01

Flow transients have been investigated in a wide range of thermal-hydraulics situations with Refrigerannt-12. Six pressures (including the reference to PWR and BWR characteristic liquid to vapour densities ratios), several periods of the flowrate transients coastdown during the simulated flow decays, and different specific mass flowrate have been studied emploiyng a circular duct test section (Dsub(i)=7,5 mm). Two heated lengths of the test section have been considered (L = 2300 and 1180 mm). Experimental data have shown the complete inadequacy of steady-state critical heat flux correlations in predicting the onset of boiling crisis during fast flow transients (half-flow decay time, tsub(h)lt5.0-6.0 s). The flow transient does not show dependence, in terms of DNB conditions ,upon the length of the test section: the ratio between transient and steady-state critical mass flowrate is not dependent on the tested geometry. The time interval from the start of the flowrate transient to the onset of DNB (time to crisis), has been experimentally determined for all the runs. Data analysis for a better theoretical prediction of the phenomenon has been accomplished, and a design correlation for DNB conditons and time to crisis prediction has been proposed

System transient response to loss of off-site power

International Nuclear Information System (INIS)

Sozer, A.

1990-01-01

A simultaneous trip of the reactor, main circulation pumps, secondary coolant pumps, and pressurizer pump due to loss of off-site power at the High Flux Isotope Reactor (HFIR) located at the Oak Ridge National Laboratory (ORNL) has been analyzed to estimate available safety margin. A computer model based on the Modular Modeling System code has been used to calculate the transient response of the system. The reactor depressurizes from 482.7 psia down to about 23 psia in about 50 seconds and remains stable thereafter. Available safety margin has been estimated in terms of the incipient boiling heat flux ratio. It is a conservative estimate due to assumed less than available primary and secondary flows and higher than normal depressurization rate. The ratio indicates no incipient boiling conditions at the hot spot. No potential damage to the fuel is likely to occur during this transient. 2 refs., 6 figs

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

Energy Technology Data Exchange (ETDEWEB)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

Generational Perspectives of Unprotected Sex and Sustainable Behavior Change in Nigeria

Directory of Open Access Journals (Sweden)

Amaechi D. Okonkwo

2013-01-01

Full Text Available Despite the HIV/AIDS pandemic and over two decades of safe-sex communication and condom social marketing in Nigeria, unmarried people continue to engage in unprotected sex. Understanding their perspectives of unprotected sex will be imperative for sustainable policy and intervention design. To realize this objective, the author synthesized Giddens’s structuration theory and Rob Stones’s structurationist project research brackets to develop a long interview guide used to elicit unmarried university students’ perspectives of influences on unprotected sex, and the feasibility of sustainable behavior change in Nigeria. Participants’ constructed unprotected sex as prescripted, and the cumulative outcome of complex institutional (structural, interpersonal, and agential influences. Their narratives challenge the popular but narrow loss of control, sensation-seeking, and ignorance theses of unprotected sex. Instead, participants’ narratives implicate an interrelated web of persuasive and insidious institutional and agential influences, in a manner that privilege neither structure nor agency. To promote safer sexual practices therefore, stakeholders must concurrently engage with institutional and agential influences on unprotected sex—and not focus on unmarried people’s sexual agencies alone, as current interventions do in Nigeria.

The effect of retarding torque during a flow transient for Tehran Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Farhadi, Kazem, E-mail: kfarhadi@aeoi.org.ir [Engineering Science Research Group, Nuclear Science Research School, Nuclear Science and Technology Research Institute, AEOI, P.O. Box 11365-3486, Tehran (Iran, Islamic Republic of)

2011-02-15

The primary cooling system of the Tehran Research Reactor (TRR) has been analysed for a possible flow transient phenomenon caused by power cut-off. All the components of the TRR primary cooling loop that offer resistance to the coolant flow are physically modelled. Differential equations of motion for the coolant in the primary piping of the TRR and for the rotating parts of the centrifugal pump are then derived. The equation of flow motion is solved simultaneously with momentum conservation equation of the rotating parts of the pump which predicts the TRR pump speed during the flow transient. Electrical and mechanical losses are measured for the TRR three-phase induction motor in order to calculate the motor retarding torque during the event. The results of the present study are compared with the other similar primary loop results. The present model shows good agreement with the existing experimental and theoretical studies.

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.

1987-01-01

The uncertainty in predicting the effectiveness of inherent shutdown (ISD) in innovative designs results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require ISD; (2) the approximation of representing all such ATWS events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. In this summary, the methodology and associated results of work used to establish probabilities of failure of inherent shutdown of innovative LMRs to the unprotected loss-of-flow (LOF) accident are discussed

FIX-II/2032, BWR Pump Trip Experiment 2032, Simulation Mass Flow and Power Transients

International Nuclear Information System (INIS)

1988-01-01

1 - Description of test facility: In the FIX-II pump trip experiments, mass flow and power transients were simulated subsequent to a total loss of power to the recirculation pumps in an internal pump boiling water reactor. The aim was to determine the initial power limit to give dryout in the fuel bundle for the specified transient. In addition, the peak cladding temperature was measured and the rewetting was studied. 2 - Description of test: Pump trip experiment 2032 was a part of test group 2, i.e. the mass flow transient was to simulate the pump coast down with a pump inertia of 11.3 kg.m -2 . The initial power in the 36-rod bundle was 4.44 MW which gave dryout after 1.4 s from the start of the flow transient. A maximum rod cladding temperature of 457 degrees C was measured. Rewetting was obtained after 7.6 s. 3 - Experimental limitations or shortcomings: No ECCS injection systems

Performance assessment of mass flow rate measurement capability in a large scale transient two-phase flow test system

International Nuclear Information System (INIS)

Nalezny, C.L.; Chapman, R.L.; Martinell, J.S.; Riordon, R.P.; Solbrig, C.W.

1979-01-01

Mass flow is an important measured variable in the Loss-of-Fluid Test (LOFT) Program. Large uncertainties in mass flow measurements in the LOFT piping during LOFT coolant experiments requires instrument testing in a transient two-phase flow loop that simulates the geometry of the LOFT piping. To satisfy this need, a transient two-phase flow loop has been designed and built. The load cell weighing system, which provides reference mass flow measurements, has been analyzed to assess its capability to provide the measurements. The analysis consisted of first performing a thermal-hydraulic analysis using RELAP4 to compute mass inventory and pressure fluctuations in the system and mass flow rate at the instrument location. RELAP4 output was used as input to a structural analysis code SAPIV which is used to determine load cell response. The computed load cell response was then smoothed and differentiated to compute mass flow rate from the system. Comparison between computed mass flow rate at the instrument location and mass flow rate from the system computed from the load cell output was used to evaluate mass flow measurement capability of the load cell weighing system. Results of the analysis indicate that the load cell weighing system will provide reference mass flows more accurately than the instruments now in LOFT

Transient burnout in flow reduction condition

International Nuclear Information System (INIS)

Iwamura, Takamichi; Kuroyanagi, Toshiyuki

1981-01-01

A transient flow reduction burnout experiment was conducted with water in a uniformly heated, vertically oriented tube. Test pressures ranged from 0.5 to 3.9 MPa. An analytical method was developed to obtain transient burnout conditions at the exit. A simple correlation to predict the deviation of the transient burnout mass velocity at the tube exit from the steady state mass velocity obtained as a function of steam-water density ratio and flow reduction rate. The correlation was also compared with the other data. (author)

Modelling and transient simulation of water flow in pipelines using WANDA Transient software

Directory of Open Access Journals (Sweden)

P.U. Akpan

2017-09-01

Full Text Available Pressure transients in conduits such as pipelines are unsteady flow conditions caused by a sudden change in the flow velocity. These conditions might cause damage to the pipelines and its fittings if the extreme pressure (high or low is experienced within the pipeline. In order to avoid this occurrence, engineers usually carry out pressure transient analysis in the hydraulic design phase of pipeline network systems. Modelling and simulation of transients in pipelines is an acceptable and cost effective method of assessing this problem and finding technical solutions. This research predicts the pressure surge for different flow conditions in two different pipeline systems using WANDA Transient simulation software. Computer models were set-up in WANDA Transient for two different systems namely; the Graze experiment (miniature system and a simple main water riser system based on some initial laboratory data and system parameters. The initial laboratory data and system parameters were used for all the simulations. Results obtained from the computer model simulations compared favourably with the experimental results at Polytropic index of 1.2.

Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-01-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1994-07-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Loss of Flow Accident (LOFA) analyses using LabView-based NRR simulator

Energy Technology Data Exchange (ETDEWEB)

Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Reactor Physics Dept.

2016-12-15

This paper presents a generic Loss of Flow Accident (LOFA) scenario module which is integrated in the LabView-based simulator to imitate a Nuclear Research Reactor (NRR) behavior for different user defined LOFA scenarios. It also provides analyses of a LOFA of a single fuel channel and its impact on operational transactions and on the behavior of the reactor. The generic LOFA scenario module includes graphs needed to clarify the effects of the LOFA under study. Furthermore, the percentage of the loss of mass flow rate, the mode of flow reduction and the start time and transient time of LOFA are user defined to add flexibility to the LOFA scenarios. The objective of integrating such generic LOFA module is to be able to deal with such incidents and avoid their significant effects. It is also useful in the development of expertise in this area and reducing the operator training and simulations costs. The results of the implemented generic LOFA module agree well with that of COBRA-IIIC code and the earlier guidebook for this series of transients.

Analysis of unprotected overcooling events in the Integral Fast Reactor

International Nuclear Information System (INIS)

Vilim, R.B.

1989-01-01

Simple analytic models are developed for predicting the response of a metal fueled, liquid-metal cooled reactor to unprotected overcooling events in the balance of plant. All overcooling initiators are shown to fall into two categories. The first category contains these events for which there is no final equilibrium state of constant overcooling, as in the case for a large steam leak. These events are analyzed using a non-flow control mass approach. The second category contains those events which will eventually equilibrate, such as a loss of feedwater heaters. A steady flow control volume analysis shows that these latter events ultimately affect the plant through the feedwater inlet to the steam generator. The models developed for analyzing these two categories provide upper bounds for the reactor's passive response to overcooling accident initiators. Calculation of these bounds for a prototypic plant indicate that failure limits -- eutectic melting, sodium boiling, fuel pin failure -- are not exceeded in any overcooling event. 2 refs

Transient characteristics of current lead losses for the large scale high-temperature superconducting rotating machine

International Nuclear Information System (INIS)

Le, T. D.; Kim, J. H.; Park, S. I.; Kim, D. J.; Kim, H. M.; Lee, H. G.; Yoon, Y. S.; Jo, Y. S.; Yoon, K. Y.

2014-01-01

To minimize most heat loss of current lead for high-temperature superconducting (HTS) rotating machine, the choice of conductor properties and lead geometry - such as length, cross section, and cooling surface area - are one of the various significant factors must be selected. Therefore, an optimal lead for large scale of HTS rotating machine has presented before. Not let up with these trends, this paper continues to improve of diminishing heat loss for HTS part according to different model. It also determines the simplification conditions for an evaluation of the main flux flow loss and eddy current loss transient characteristics during charging and discharging period.

SASSYS validation with the EBR-II shutdown heat removal tests

International Nuclear Information System (INIS)

Herzog, J.P.

1989-01-01

SASSYS is a coupled neutronic and thermal hydraulic code developed for the analysis of transients in liquid metal cooled reactors (LMRs). The code is especially suited for evaluating of normal reactor transients -- protected (design basis) and unprotected (anticipated transient without scram) transients. Because SASSYS is heavily used in support of the IFR concept and of innovative LMR designs, such as PRISM, a strong validation base for the code must exist. Part of the validation process for SASSYS is analysis of experiments performed on operating reactors, such as the metal fueled Experimental Breeder Reactor -- II (EBR-II). During the course of a series of historic whole-plant experiments, EBR-II illustrated key safety features of metal fueled LMRs. These experiments, the Shutdown Heat Removal Tests (SHRT), culminated in unprotected loss of flow and loss of heat sink transients from full power and flow. Analysis of these and earlier SHRT experiments constitutes a vital part of SASSYS validation, because it facilitates scrutiny of specific SASSYS models and of integrated code capability. 12 refs., 11 figs

Inherently safe shutdown of a liquid metal reactor upon a loss of intermediate cooling

International Nuclear Information System (INIS)

Feldman, E.E.; Mohr, D.

1986-01-01

Two unprotected (i.e., no scram or plant protection system action) loss-of-heat sink transients are scheduled to be performed on the Experimental Breeder Reactor-II in the Spring of 1986. One is to be initiated from full power (about 60 MW) and the other from half power. The loss-of-heat sink in each test is to be accomplished by essentially stopping the secondary-loop sodium coolant flow in about 20 s. Pretest predictions for the two tests, provided herein, show the reactor to passively shut down with average reactor outlet coolant temperatures rising no more that 5/degree/C before dropping and asymptotically approaching a quenching (or ''smothering'') temperature. This mild behavior is seen to be a direct consequence of the substantial negative reactivity feedback of the reactor coupled with the heat capacitance of the massive sodium pool in the primary tank. 14 refs., 11 figs., 1 tab

MINET: transient analysis of fluid-flow and heat-transfer networks

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Guppy, J.G.; Nepsee, T.C.

1983-01-01

MINET, a computer code developed for the steady-state and transient analysis of fluid-flow and heat-transfer networks, is described. The code is based on a momentum integral network method, which offers significant computational advantages in the analysis of large systems, such as the balance of plant in a power-generating facility. An application is discussed in which MINET is coupled to the Super System Code (SSC), an advanced generic code for the transient analysis of loop- or pool-type LMFBR systems. In this application, the ability of the Clinch River Breeder Reactor Plant to operate in a natural circulation mode following an assumed loss of all electric power, was assessed. Results from the MINET portion of the calculations are compared against those generated independently by the Clinch River Project, using the DEMO code

Effect of load transients on SOFC operation—current reversal on loss of load

Science.gov (United States)

Gemmen, Randall S.; Johnson, Christopher D.

The dynamics of solid oxide fuel cell (SOFC) operation have been considered previously, but mainly through the use of one-dimensional codes applied to co-flow fuel cell systems. In this paper several geometries are considered, including cross-flow, co-flow, and counter-flow. The details of the model are provided, and the model is compared with some initial experimental data. For parameters typical of SOFC operation, a variety of transient cases are investigated, including representative load increase and decrease and system shutdown. Of particular note for large load decrease conditions (e.g., shutdown) is the occurrence of reverse current over significant portions of the cell, starting from the moment of load loss up to the point where equilibrated conditions again provide positive current. Consideration is given as to when such reverse current conditions might most significantly impact the reliability of the cell.

Large scale applicability of a Fully Adaptive Non-Intrusive Spectral Projection technique: Sensitivity and uncertainty analysis of a transient

International Nuclear Information System (INIS)

Perkó, Zoltán; Lathouwers, Danny; Kloosterman, Jan Leen; Hagen, Tim van der

2014-01-01

Highlights: • Grid and basis adaptive Polynomial Chaos techniques are presented for S and U analysis. • Dimensionality reduction and incremental polynomial order reduce computational costs. • An unprotected loss of flow transient is investigated in a Gas Cooled Fast Reactor. • S and U analysis is performed with MC and adaptive PC methods, for 42 input parameters. • PC accurately estimates means, variances, PDFs, sensitivities and uncertainties. - Abstract: Since the early years of reactor physics the most prominent sensitivity and uncertainty (S and U) analysis methods in the nuclear community have been adjoint based techniques. While these are very effective for pure neutronics problems due to the linearity of the transport equation, they become complicated when coupled non-linear systems are involved. With the continuous increase in computational power such complicated multi-physics problems are becoming progressively tractable, hence affordable and easily applicable S and U analysis tools also have to be developed in parallel. For reactor physics problems for which adjoint methods are prohibitive Polynomial Chaos (PC) techniques offer an attractive alternative to traditional random sampling based approaches. At TU Delft such PC methods have been studied for a number of years and this paper presents a large scale application of our Fully Adaptive Non-Intrusive Spectral Projection (FANISP) algorithm for performing the sensitivity and uncertainty analysis of a Gas Cooled Fast Reactor (GFR) Unprotected Loss Of Flow (ULOF) transient. The transient was simulated using the Cathare 2 code system and a fully detailed model of the GFR2400 reactor design that was investigated in the European FP7 GoFastR project. Several sources of uncertainty were taken into account amounting to an unusually high number of stochastic input parameters (42) and numerous output quantities were investigated. The results show consistently good performance of the applied adaptive PC

Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

International Nuclear Information System (INIS)

Chenu, A.

2011-10-01

liquid film after dryout onset. The validation of the extended TRACE code has been achieved through the successful simulation of out-of-pile experiments. A review of available sodium boiling test data has first been carried out, and complementary tests have been selected to assess the quality of the different physical models like the pressure drop and cooling limits under quasi steady-state conditions, as well as the simulation of a loss-of-flow transient. The extended TRACE code has demonstrated its capacity to predict the main thermal-hydraulics characteristics such as the single- and two-phase pressure drop and heat transfer, as also the boiling inception, void fraction evolution and expansion of the boiling region, pressure evolution, as well as coolant and clad temperatures. The natural convection test conducted in 2009 in the Phenix reactor has been used to validate TRACE single-phase sodium flow modelling. Data from the Phenix test have been used for the validation of the FAST code system. Analyses based on a point-kinetics TRACE model and on coupled TRACE/PARCS 3D-kinetics modelling have enabled an in-depth understanding of the transient behaviour of a sodium-cooled fast reactor core, leading to potential improvements in the FAST code system. The experimental power evolution could be satisfactorily reproduced within the measurement uncertainties with both models, and the detailed analysis of the core neutronics has enabled one to define the most important reactivity feedbacks taking place during the considered transient. The developed tool has been applied to the simulation of a hypothetical, unprotected loss-of-flow event for one of the European SFR core concepts. This study has demonstrated the new calculation tool’s capability to adequately simulate the core response through the modelling of single- and two-phase sodium flow, coupled to 3D neutron kinetics. Thereby, the space-dependent reactivity feedbacks, such as the void and Doppler effects, have been

Preliminary Evaluation of the Diverse Protection System in PGSFR

International Nuclear Information System (INIS)

Jeong, Taekyeong; Chang, Won-Pyo; Seong, Seung Hwan; Ahn, Sang June; Kang, Seok Hun; Choi, Chiwoong; Yoo, Jin; Lee, Kwi Lim; Lee, Seung Won; Jeong, Jae-Ho; Ha, Kwi-Seok

2015-01-01

The anticipated transient without scram (ATWS) is defined as an abnormal transient with failure of scram actuation. It is one of the “worst case” accident based on the United States Nuclear Regulatory Commission (U.S.NRC). Consideration frequently motivates the NRC to take regulatory action. An evaluation of this event is also a general requirement due to a potential safety issue that may lead to core damage under postulated condition. This paper estimated the set-points sensitivity test of the diverse protection system (DPS) related with unprotected events of the prototype generation-IV sodium cooled fast reactor (PGSFR) including unprotected transient over power (UTOP) and unprotected loss of flow (ULOF) by MARS-LMR code. The variation of the power to flow (P/Q) of UTOP and ULOF is illustrated to conduct the set-points sensitivity test of DPS. Also we estimated the effect of the DPS introduction after selecting UTOP, ULOF event as the unprotected events which are predicted to aggravate the events. This paper estimated the set-points sensitivity test of DPS related with unprotected events of PGSFR including UTOP and ULOF by MARS-LMR code. The results indicated that there is no significant difference in both RPS and DPS about the initiating time of each event. Therefore, this study found that the urgent manage for safety of the reactor when RPS failed is possible by the applying DPS

Preliminary Evaluation of the Diverse Protection System in PGSFR

Energy Technology Data Exchange (ETDEWEB)

Jeong, Taekyeong; Chang, Won-Pyo; Seong, Seung Hwan; Ahn, Sang June; Kang, Seok Hun; Choi, Chiwoong; Yoo, Jin; Lee, Kwi Lim; Lee, Seung Won; Jeong, Jae-Ho; Ha, Kwi-Seok [KAERI, Daejeon (Korea, Republic of)

2015-05-15

The anticipated transient without scram (ATWS) is defined as an abnormal transient with failure of scram actuation. It is one of the “worst case” accident based on the United States Nuclear Regulatory Commission (U.S.NRC). Consideration frequently motivates the NRC to take regulatory action. An evaluation of this event is also a general requirement due to a potential safety issue that may lead to core damage under postulated condition. This paper estimated the set-points sensitivity test of the diverse protection system (DPS) related with unprotected events of the prototype generation-IV sodium cooled fast reactor (PGSFR) including unprotected transient over power (UTOP) and unprotected loss of flow (ULOF) by MARS-LMR code. The variation of the power to flow (P/Q) of UTOP and ULOF is illustrated to conduct the set-points sensitivity test of DPS. Also we estimated the effect of the DPS introduction after selecting UTOP, ULOF event as the unprotected events which are predicted to aggravate the events. This paper estimated the set-points sensitivity test of DPS related with unprotected events of PGSFR including UTOP and ULOF by MARS-LMR code. The results indicated that there is no significant difference in both RPS and DPS about the initiating time of each event. Therefore, this study found that the urgent manage for safety of the reactor when RPS failed is possible by the applying DPS.

Summary and evaluation: fuel dynamics loss-of-flow experiments (tests L2, L3, and L4)

International Nuclear Information System (INIS)

Barts, E.W.; Deitrich, L.W.; Eberhart, J.G.; Fischer, A.K.; Meek, C.C.

1975-09-01

Three similar experiments conducted to support the analyses of hypothetical LMFBR unprotected-loss-of-flow accidents are summarized and evaluated. The tests, designated L2, L3, and L4, provided experimental data against which accident-analysis codes could be compared, so as to guide further analysis and modeling of the initiating phases of the hypothetical accident. The tests were conducted using seven-pin bundles of mixed-oxide fuel pins in Mark-II flowing-sodium loops in the TREAT reactor. Test L2 used fresh fuel. Tests L3 and L4 used irradiated fuel pins having, respectively, ''intermediate-power'' (no central void) and ''high-power'' (fully developed central void) microstructure. 12 references

Study of transient burnout under flow reduction condition

International Nuclear Information System (INIS)

Iwamura, Takamichi

1986-09-01

Transient burnout characteristics of a fuel rod under a rapid flow reduction condition of a light water reactor were experimentally and analytically studied. The test sections were uniformly heated vertical tube and annulus with the heated length of 800 mm. Test pressures ranged 0.5 ∼ 3.9 MPa, heat fluxes 2,160 ∼ 3,860 KW/m 2 , and flow reduction rates 0.44 ∼ 770 %/s. The local flow condition during flow reduction transients were calculated with a separate flow model. The two-fluid/three-field thermal-hydraulic code, COBRA/TRAC, was also used to investigate the liquid film behavior on the heated surface. The major results obtained in the present study are as follows: The onset of burnout under a rapid flow reduction condition was caused by a liquid film dryout on the heated surface. With increasing flow reduction rate beyond a threshold, the burnout mass velocity at the inlet became lower than the steady-state burnout mass velocity. This is explained by the fact that the vapor flow rate continues to increase due to the delay of boiling boundary movement and the resultant high vapor velocity sustains the liquid film flow after the inlet flow rate reaches the steady-state burnout flow rate. The ratio of inlet burnout mass velocities between flow reduction transient and steady-state became smaller with increasing system pressure because of the lower vapor velocity due to the lower vapor specific volume. Flow reduction burnout occurred when the outlet quality agreed with the steady-state burnout quality within 10 %, suggesting that the local condition burnout model can be used for flow reduction transients. Based on this model, a method to predict the time to burnout under a flow reduction condition in a uniformly heated tube was developed. The calculated times to burnout agreed well with some experimental results obtained by the Author, Cumo et al., and Moxon et al. (author)

Analysis of forced convective transient boiling by homogeneous model of two-phase flow

International Nuclear Information System (INIS)

Kataoka, Isao

1985-01-01

Transient forced convective boiling is of practical importance in relation to the accident analysis of nuclear reactor etc. For large length-to-diameter ratio, the transient boiling characteristics are predicted by transient two-phase flow calculations. Based on homogeneous model of two-phase flow, the transient forced convective boiling for power and flow transients are analysed. Analytical expressions of various parameters of transient two-phase flow have been obtained for several simple cases of power and flow transients. Based on these results, heat flux, velocity and time at transient CHF condition are predicted analytically for step and exponential power increases, and step, exponential and linear velocity decreases. The effects of various parameters on heat flux, velocity and time at transient CHF condition have been clarified. Numerical approach combined with analytical method is proposed for more complicated cases. Solution method for pressure transient are also described. (author)

Transient flow assurance for determination of operational control of heavy oil pipelines

Energy Technology Data Exchange (ETDEWEB)

Cabrejo, Victor [TransCanada Pipelines Ltd, (Canada); Mohitpour, Mo [Tempsys Pipeline Solutions Inc., (Canada)

2010-07-01

Pipeline transmission systems have been designed traditionally using steady state simulations. Steady state simulation provided sufficient values for simple systems, but is limited in dealing with surges in flow rates, loss of facilities and facility operation. A dynamic approach is required to test the capacity of a system for various fluids. This paper investigated the use of transient analysis of liquid pipelines in order to improve the design of these pipelines and to achieve operational benefits. The transient method and its use are discussed. Dynamic analysis was applied to the Keystone Pipeline Project. The purpose of the study was first to determine the system capacity and data for transportation of Heavy DilBit, and then to implement batch transportation of a volume of synthetic crude oil. It was found that the use of transient modeling in design and operational assessment of a liquid pipeline ensures system capability, control, safety and integrity.

Analysis of the transient compressible vapor flow in heat pipes

Science.gov (United States)

Jang, J. H.; Faghri, A.; Chang, W. S.

1989-01-01

The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures.

Analysis of the transient compressible vapor flow in heat pipe

International Nuclear Information System (INIS)

Jang, J.H.; Faghri, A.; Chang, W.S.

1989-07-01

The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures

Analysis of the transient compressible vapor flow in heat pipe

Science.gov (United States)

Jang, Jong Hoon; Faghri, Amir; Chang, Won Soon

1989-01-01

The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures.

Transient flow analysis of integrated valve opening process

Energy Technology Data Exchange (ETDEWEB)

Sun, Xinming; Qin, Benke; Bo, Hanliang, E-mail: bohl@tsinghua.edu.cn; Xu, Xingxing

2017-03-15

Highlights: • The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology and the integrated valve (IV) is the key control component. • The transient flow experiment induced by IV is conducted and the test results are analyzed to get its working mechanism. • The theoretical model of IV opening process is established and applied to get the changing rule of the transient flow characteristic parameters. - Abstract: The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology and the IV is the key control component. The working principle of integrated valve (IV) is analyzed and the IV hydraulic experiment is conducted. There is transient flow phenomenon in the valve opening process. The theoretical model of IV opening process is established by the loop system control equations and boundary conditions. The valve opening boundary condition equation is established based on the IV three dimensional flow field analysis results and the dynamic analysis of the valve core movement. The model calculation results are in good agreement with the experimental results. On this basis, the model is used to analyze the transient flow under high temperature condition. The peak pressure head is consistent with the one under room temperature and the pressure fluctuation period is longer than the one under room temperature. Furthermore, the changing rule of pressure transients with the fluid and loop structure parameters is analyzed. The peak pressure increases with the flow rate and the peak pressure decreases with the increase of the valve opening time. The pressure fluctuation period increases with the loop pipe length and the fluctuation amplitude remains largely unchanged under different equilibrium pressure conditions. The research results lay the base for the vibration reduction analysis of the CRHDS.

Core size effects on safety performances of LMRs

Energy Technology Data Exchange (ETDEWEB)

Na, Byung Chan; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). In the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow (ULOF) and the unprotected transient overpower (UTOP). Margins to fuel melting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events. 6 refs., 4 figs., 2 tabs. (Author)

Core size effects on safety performances of LMRs

Energy Technology Data Exchange (ETDEWEB)

Na, Byung Chan; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). In the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow (ULOF) and the unprotected transient overpower (UTOP). Margins to fuel melting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events. 6 refs., 4 figs., 2 tabs. (Author)

Measurements of flow-rate transients in one-phase liquid flow

International Nuclear Information System (INIS)

Mueller-Roos, J.

1975-01-01

A report is given on a method to determine flow-rate transients in a one-phase flow. Periodic temperature signals are superposed on the flow, from which flow times are calculated through correlation each over a half period. The evaluation is carried out according to the digitalization 'off-line' on a large computer. Rate peaks of over 100% within 1.9 s were qualitatively and quantitatively well represented. (orig./LH) [de

Kinetic Energy Losses and Efficiency of an Axial Turbine Stage in Numerical Modeling of Unsteady Flows

Directory of Open Access Journals (Sweden)

A. S. Laskin

2015-01-01

Full Text Available The article presents the results of numerical investigation of kinetic energy (KE loss and blading efficiency of the single-stage axial turbine under different operating conditions, characterized by the ratio u/C0. The calculations are performed by stationary (Stage method and nonstationary (Transient method methods using ANSYS CFX. The novelty of this work lies in the fact that the numerical simulation of steady and unsteady flows in a turbine stage is conducted, and the results are obtained to determine the loss of KE, both separately by the elements of the flow range and their total values, in the stage efficiency as well. The results obtained are compared with the calculated efficiency according to one-dimensional theory.To solve these problems was selected model of axial turbine stage with D/l = 13, blade profiles of rotor and stator of constant cross-section, similar to tested ones in inverted turbine when = 0.3. The degree of reactivity ρ = 0.27, the rotor speed was varied within the range 1000 ÷ 1800 rev/min.Results obtained allow us to draw the following conclusions:1. The level of averaged coefficients of total KE losses in the range of from 0.48 to 0.75 is from 18% to 21% when calculating by the Stage method and from 21% to 25% by the Transient one.2. The level of averaged coefficients of KE losses with the output speed of in the specified range is from 9% to 13%, and almost the same when in calculating by Stage and Transient methods.3. Levels of averaged coefficients of KE loss in blade tips (relative to the differential enthalpies per stage are changed in the range: from 4% to 3% (Stage and are stored to be equal to 5% (Transient; from 5% to 6% (Stage and from 6% to 8% (Transient.4. Coefficients of KE losses in blade tips GV and RB are higher in calculations of the model stage using the Transient method than the Stage one, respectively, by = 1.5 ÷ 2.5% and = 4 ÷ 5% of the absolute values. These are values to characterize the KE

Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

International Nuclear Information System (INIS)

Schikorr, W.M.

2001-01-01

The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k eff ∼0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k eff ∼0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early

Transient response in granular bounded heap flows

Science.gov (United States)

Xiao, Hongyi; Ottino, Julio M.; Lueptow, Richard M.; Umbanhowar, Paul B.

2017-11-01

Heap formation, a canonical granular flow, is common in industry and is also found in nature. Here, we study the transition between steady flow states in quasi-2D bounded heaps by suddenly changing the feed rate from one fixed value to another. During the transition, in both experiments and discrete element method simulations, an additional wedge of flowing particles propagates over the rising free surface. The downstream edge of the wedge - the wedge front - moves downstream with velocity inversely proportional to the square root of time. An additional longer duration transient process continues after the wedge front reaches the downstream wall. The transient flux profile during the entire transition is well modeled by a diffusion-like equation derived from local mass balance and a local linear relation between the flux and the surface slope. Scalings for the transient kinematics during the flow transitions are developed based on the flux profiles. Funded by NSF Grant CBET-1511450.

Accidents of loss of flow for the ETTR-2 reactor; deterministic analysis

International Nuclear Information System (INIS)

El-Messiry, A.M.

2000-01-01

The main objective for reactor safety is to keep the fuel in a thermally safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accident (DBA) as loss of flow accident (LOFA), is required assessing reactor safety. The present work concerns this transients applied to Egypt Test and Research Reactor ETRR-3 (new reactor). An accident analysis code FLOWTR is developed to investigate the thermal behaviour of the core during such flow transients. The active core is simulated by two channels: 1 - hot channel (HC), and 2 - average channel (AC) representing the remainder of the core. Each channel is divided into four axial sections. The external loop, core plenums, and core chimney are simulated by different dynamic loops. The code includes modules for pump cast down, flow regimes, decay heat, temperature distributions, and feedback coefficients. FLOWTR is verified against results from RETRAN code, THERMIC code and commissioning tests for null transient case. The comparison shows a good agreement. The study indicates that for LOFA transients, provided the scram system is available, the core is shutdown safely by low flow signal (496.6 kg/s) at 1.4 s were the HC temperature reaches the maximum value, 45.64 o C after shutdown. On the other hand, if the scram system is unavailable, and at t = 47.33 s, the core flow decreases to 67.41 kg/s, the HC temperature increases to 164.02 o C, and the HC clad surface heat flux exceeds its critical value of 400.00 W/cm 2 resulting of fuel burnout. (author)

Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

International Nuclear Information System (INIS)

Cahalan, J.; Wigeland, R.; Friedel, G.; Kussmaul, G.; Royl, P.; Moreau, J.; Perks, M.

1990-01-01

In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs

Transient analysis of a variable speed rotary compressor

International Nuclear Information System (INIS)

Park, Youn Cheol

2010-01-01

A transient simulation model of a rolling piston type rotary compressor is developed to predict the dynamic characteristics of a variable speed compressor. The model is based on the principles of conservation, real gas equations, kinematics of the crankshaft and roller, mass flow loss due to leakage, and heat transfer. For the computer simulation of the compressor, the experimental data were obtained from motor performance tests at various operating frequencies. Using the developed model, re-expansion loss, friction loss, mass flow loss and heat transfer loss is estimated as a function of the crankshaft speed in a variable speed compressor. In addition, the compressor efficiency and energy losses are predicted at various compressor-operating frequencies. Since the transient state of the compressor strongly depends on the system, the developed model is combined with a transient system simulation program to get transient variations of the compression process in the system. Motor efficiency, mechanical efficiency, motor torque and volumetric efficiency are calculated with respect to variation of the driving frequency in a rotary compressor.

Study of transient burnout characteristics under flow reduction condition

International Nuclear Information System (INIS)

Iwamura, Takamichi; Kuroyanagi, Toshiyuki

1984-03-01

As part of a study of the thermal behavior of fuel rods during Power-Cooling-Mismatch (PCM) accidents in light water reactors, burnout characteristics in a uniformly heated, vertically oriented tube or annulus, under flow reduction condition, were experimentally studied. Test pressures ranged 0.1--3.9 MPa and flow reduction rates 0.44--1100%/s. An analytical method is developed to obtain the local mass velocity during a transient condition. The major results are as follows: With increasing flow reduction rate beyond a threshold, transient burnout mass velocity at the inlet was lower than that in steady state tests under the experimental pressures. The higher system pressure resulted in the less transient effects. At pressures higher than 2.0 MPa and flow reduction rates lower than 20%/s, the local burnout mass velocity agreed with the steady state burnout mass velocity, whereas the local burnout mass velocity became higher than the steady state burnout mass velocity at flow reduction rates higher than 20%/s. At pressures lower than 1 MPa, with increasing flow reduction rate beyond the threshold value of 2%/s, the local burnout mass velocity was lower than the steady state burnout mass velocity. An empirical correlation is presented to give the ratio of the transient to the steady state burnout mass velocities at the burnout location as a function of the steam-water density ratio and the flow reduction rate. The experimental results by Cumo et al. agree with the correlation. The correlation, however, cannot predict the experimental results at higher flow reduction rates beyond 40%/s. (author)

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

International Nuclear Information System (INIS)

Mendler, O.J.; Takeuchi, K.; Young, M.Y.

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

Energy Technology Data Exchange (ETDEWEB)

Mendler, O J; Takeuchi, K; Young, M Y

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

Transient flow characteristics of a high speed rotary valve

Science.gov (United States)

Browning, Patrick H.

were experimentally mapped as a function of valve speed, inter-cylinder pressure ratios and volume ratios and the results were compared to compressible flow theoretical models. Specifically, the transient behavior suggested a short-lived loss-mode initiation closely resembled by shock tube theory followed by a quasi-steady flow regime resembling choked flow behavior. An empirical model was then employed to determine the useful range of the CCV design as applied to a four-stroke CIBAI engine cycle modeled using a 1-D quasi-steady numerical method, with particular emphasis on the cyclic timing of the CCV opening. Finally, a brief discussion of a high-temperature version of the CCV design is presented.

Present status of numerical analysis on transient two-phase flow

International Nuclear Information System (INIS)

Akimoto, Masayuki; Hirano, Masashi; Nariai, Hideki.

1987-01-01

The Special Committee for Numerical Analysis of Thermal Flow has recently been established under the Japan Atomic Energy Association. Here, some methods currently used for numerical analysis of transient two-phase flow are described citing some information given in the first report of the above-mentioned committee. Many analytical models for transient two-phase flow have been proposed, each of which is designed to describe a flow by using differential equations associated with conservation of mass, momentum and energy in a continuous two-phase flow system together with constructive equations that represent transportation of mass, momentum and energy though a gas-liquid interface or between a liquid flow and the channel wall. The author has developed an analysis code, called MINCS, that serves for systematic examination of conservation equation and constructive equations for two-phase flow models. A one-dimensional, non-equilibrium two-liquid flow model that is used as the basic model for the code is described. Actual procedures for numerical analysis is shown and some problems concerning transient two-phase analysis are described. (Nogami, K.)

Accidents and transients analyses of a super fast reactor with single flow pass core

International Nuclear Information System (INIS)

Sutanto,; Oka, Yoshiaki

2014-01-01

Highlights: • Safety analysis of a Super FR with single flow pass core is conducted. • Loss of feed water flow leads to a direct effect on the loss of fuel channel flow. • The core pressure is sensitive to LOCA accidents due to the direct effect. • Small LOCA introduces a critical break. • The safety criteria for all selected events are satisfied. - Abstract: The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST

The Transient Elliptic Flow of Power-Law Fluid in Fractal Porous Media

Institute of Scientific and Technical Information of China (English)

宋付权; 刘慈群

2002-01-01

The steady oil production and pressure distribution formulae of vertically fractured well for power-law non-Newtonian fluid were derived on the basis of the elliptic flow model in fractal reservoirs. The corresponding transient flow in fractal reservoirs was studied by numerical differentiation method: the influence of fractal index to transient pressure of vertically fractured well was analyzed. Finally the approximate analytical solution of transient flow was given by average mass conservation law. The study shows that using elliptic flow method to analyze the flow of vertically fractured well is a simple method.

Transient compressible flows in porous media

International Nuclear Information System (INIS)

Morrison, F.A. Jr.

1975-09-01

Transient compressible flow in porous media was investigated analytically. The major portion of the investigation was directed toward improving and understanding of dispersion in these flows and developing rapid accurate numerical techniques for predicting the extent of dispersion. The results are of interest in the containment of underground nuclear experiments. The transient one-dimensional transport of a trace component in a gas flow is analyzed. A conservation equation accounting for the effects of convective transport, dispersive transport, and decay, is developed. This relation, as well as a relation governing the fluid flow, is used to predict trace component concentration as a function of position and time. A detailed analysis of transport associated with the isothermal flow of an ideal gas is done. Because the governing equations are nonlinear, numerical calculations are performed. The ideal gas flow is calculated using a highly stable implicit iterative procedure with an Eulerian mesh. In order to avoid problems of anomolous dispersion associated with finite difference calculation, trace component convection and dispersion are calculated using a Lagrangian mesh. Details of the Eulerian-Lagrangian numerical technique are presented. Computer codes have been developed and implemented on the Lawrence Livermore Laboratory computer system

Analysis of the one-dimensional transient compressible vapor flow in heat pipes

Science.gov (United States)

Jang, Jong H.; Faghri, Amir; Chang, Won S.

1991-01-01

The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds as well as high mass flow rates are successfully predicted.

Application of a film flow model to predicting burnout under transient conditions

International Nuclear Information System (INIS)

Leslie, D.C.; Kirby, G.J.

1967-08-01

The film flow model developed previously has been generalised to transient situations by assuming that only convection is changed by the transient; evaporation, deposition and entrainment are assumed to be unaffected. A computer code TRABUT computes the time behaviour of the mass velocity and the quality by the method of characteristics, and then integrates the film flow equations along the same characteristics until the point of burn-out or zero film flow is reached. The time delay between the onset of a transient and burn-out has been computed both for flux and flow transients. These computations have been compared with those made using the standard local conditions hypothesis. The film flow model gives shorter delays in almost all cases, but the difference would not be detectable with present experimental techniques. (author)

OPR1000 RCP Flow Coastdown Analysis using SPACE Code

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.

Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

2010-07-01

The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

One-dimensional transient unequal velocity two-phase flow by the method of characteristics

International Nuclear Information System (INIS)

Rasouli, F.

1981-01-01

An understanding of two-phase flow is important when one is analyzing the accidental loss of coolant or when analyzing industrial processes. If a pipe in the steam generator of a nuclear reactor breaks, the flow will remain critical (or choked) for almost the entire blowdown. For this reason the knowledge of the two-phase maximum (critical) flow rate is important. A six-equation model--consisting of two continuity equations, two energy equations, a mixture momentum equation, and a constitutive relative velocity equation--is solved numerically by the method of characteristics for one-dimensional, transient, two-phase flow systems. The analysis is also extended to the special case of transient critical flow. The six-equation model is used to study the flow of a nonequilibrium sodium-argon system in a horizontal tube in which the nonequilibrium sodium-argon system in a horizontal tube in which the critical flow condition is at the entrance. A four-equation model is used to study the pressure-pulse propagation rate in an isothermal air-water system, and the results that are found are compared with the experimental data. Proper initial and boundary conditions are obtained for the blowdown problem. The energy and mass exchange relations are evaluated by comparing the model predictions with results of void-fraction and heat-transfer experiments. A simplified two-equation model is obtained for the special case of two incompressible phases. This model is used in the preliminary analysis of batch sedimentation. It is also used to predict the shock formation in the gas-solid fluidized bed

New transient-flow modelling of a multiple-fractured horizontal well

International Nuclear Information System (INIS)

Jia, Yong-Lu; Wang, Ben-Cheng; Nie, Ren-Shi; Wang, Dan-Ling

2014-01-01

A new transient-flow modelling of a multiple-fractured horizontal well is presented. Compared to conventional modelling, the new modelling considered more practical physical conditions, such as various inclined angles for different fractures, different fracture intervals, different fracture lengths and partially penetrating fractures to formation. A kind of new mathematical method, including a three-dimensional eigenvalue and orthogonal transform, was created to deduce the exact analytical solutions of pressure transients for constant-rate production in real space. In order to consider a wellbore storage coefficient and skin factor, we used a Laplace-transform approach to convert the exact analytical solutions to the solutions in Laplace space. Then the numerical solutions of pressure transients in real space were gained using a Stehfest numerical inversion. Standard type curves were plotted to describe the transient-flow characteristics. Flow regimes were clearly identified from type curves. Furthermore, the differences between the new modelling and the conventional modelling in pressure transients were especially compared and discussed. Finally, an example application to show the accordance of the new modelling with real conditions was implemented. Our new modelling is different from, but more practical than, conventional modelling. (paper)

THEBES: a thermal hydraulic code for the calculation of transient two phase flow in bundle geometry

International Nuclear Information System (INIS)

Camous, F.

1983-01-01

The three dimensional thermal hydraulic code THEBES, capable to calculate transient boiling of sodium in rod bundles is described here. THEBES, derived from the transient single phase code SABRE-2A, was developed in CADARACHE by the SIES to analyse the SCARABEE N loss of flow experiments. This paper also presents the results of tests which were performed against various types of experiments: (1) transient boiling in a 7 pin bundle simulating a partial blockage at the bottom of a subassembly (rapid transient SCARABEE 7.2 experiment), (2) transient boiling in a 7 pin bundle simulating a coolant coast down (slow transient SCARABEE 7.3 experiment), (3) steady local and generalised boiling in a 19 pin bundle (GR 19 I experiment), (4) transient boiling in a 19 pin bundle simulating a coolant coast down (GR 19 I experiment), (5) steady local boiling in a 37 pin bundle with internal blockage (MOL 7C experiment). Excellent agreement was found between calculated and experimental results for these different situations. Our conclusion is that THEBES is able to calculate transient boiling of sodium in rod bundles in a quite satisfying way

Isolated transient loss of consciousness is an indicator of significant injury.

Science.gov (United States)

Owings, J T; Wisner, D H; Battistella, F D; Perlstein, J; Walby, W F; Tharratt, R S

1998-09-01

To determine if isolated transient loss of consciousness is an indicator of significant injury. University-based level I trauma center. Phase 1 retrospective case series of all patients with trauma admitted directly from the emergency department to the operating room or an intensive care unit who had transient loss of consciousness as their only trauma triage criterion. Phase 2 prospective case series of all trauma patients transported by emergency medical system personnel with transient loss of consciousness as their only trauma triage criterion. Emergency operation and intensive care unit admission. Phase 1: From January 1, 1992, to March 31, 1995, we admitted 10255 patients with trauma. Three hundred seven (3%) met the enrollment criteria and were admitted to the operating room (n = 168) or intensive care unit (n = 139). Of these, 58 (18.9%) were taken to the operating room emergently to manage life-threatening injuries: 11 (4%) had craniotomies and 47 (15%) had non-neurosurgical operations. Phase 2: From July 1 to December 31, 1996, 2770 trauma patients were transported to our facility; 135 (4.9%) met the enrollment criteria. Forty-one (30.4%) of these required admission, and 6 (4.4%) were taken emergently to the operating room from the emergency department (1 [1%] for a craniotomy, 3 [2.2%] for intra-abdominal bleeding, and 2 [1.5%] for other procedures). Two (1.5%) of the 135 patients died. Patients with isolated transient loss of consciousness are at significant risk of critical surgical and neurosurgical injuries. These patients should be triaged to trauma centers or hospitals with adequate imaging, surgical, and neurosurgical resources.

Simulation of non-isothermal transient flow in gas pipeline

Energy Technology Data Exchange (ETDEWEB)

Ferreira Junior, Luis Carlos; Soares, Matheus; Lima, Enrique Luis; Pinto, Jose Carlos [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Quimica; Muniz, Cyro; Pires, Clarissa Cortes; Rochocz, Geraldo [ChemTech, Rio de Janeiro, RJ (Brazil)

2009-07-01

Modeling of gas pipeline usually considers that the gas flow is isothermal (or adiabatic) and that pressure changes occur instantaneously (quasi steady state approach). However, these assumptions are not valid in many important transient applications (changes of inlet and outlet flows/pressures, starting and stopping of compressors, changes of controller set points, among others). Besides, the gas properties are likely to depend simultaneously on the pipe position and on the operation time. For this reason, a mathematical model is presented and implemented in this paper in order to describe the gas flow in pipeline when pressure and temperature transients cannot be neglected. The model is used afterwards as a tool for reconciliation of available measured data. (author)

Antibiotic losses from unprotected manure stockpiles.

Science.gov (United States)

Dolliver, Holly A S; Gupta, Satish C

2008-01-01

Manure management is a major concern in livestock production systems. Although historically the primary concerns have been nutrients and pathogens, manure is also a source of emerging contaminants, such as antibiotics, to the environment. There is a growing concern that antibiotics in manure are reaching surface and ground waters and contributing to the development and spread of antibiotic resistance in the environment. One such pathway is through leaching and runoff from manure stockpiles. In this study, we quantified chlortetracycline, monensin, and tylosin losses in runoff from beef manure stockpiles during two separate but consecutive experiments representing different weather conditions (i.e., temperature and precipitation amount and form). Concentrations of chlortetracycline, monensin, and tylosin in runoff were positively correlated with initial concentrations of antibiotics in manure. The highest concentrations of chlortetracycline, monensin, and tylosin in runoff were 210, 3175, and 2544 microg L(-1), respectively. Relative antibiotic losses were primarily a function of water losses. In the experiment that had higher runoff water losses, antibiotic losses ranged from 1.2 to 1.8% of total extractable antibiotics in manure. In the experiment with lower runoff water losses, antibiotic losses varied from 0.2 to 0.6% of the total extractable antibiotics in manure. Manure analysis over time suggests that in situ degradation is an important mechanism for antibiotic losses. Degradation losses during manure stockpiling may exceed cumulative losses from runoff events. Storing manure in protected (i.e., covered) facilities could reduce the risk of aquatic contamination associated with manure stockpiling and other outdoor manure management practices.

CONNECTING FLARES AND TRANSIENT MASS-LOSS EVENTS IN MAGNETICALLY ACTIVE STARS

Energy Technology Data Exchange (ETDEWEB)

Osten, Rachel A. [Space Telescope Science Institute 3700 San Martin Drive, Baltimore, MD 21218 (United States); Wolk, Scott J., E-mail: osten@stsci.edu [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge MA 02138 (United States)

2015-08-10

We explore the ramification of associating the energetics of extreme magnetic reconnection events with transient mass-loss in a stellar analogy with solar eruptive events. We establish energy partitions relative to the total bolometric radiated flare energy for different observed components of stellar flares and show that there is rough agreement for these values with solar flares. We apply an equipartition between the bolometric radiated flare energy and kinetic energy in an accompanying mass ejection, seen in solar eruptive events and expected from reconnection. This allows an integrated flare rate in a particular waveband to be used to estimate the amount of associated transient mass-loss. This approach is supported by a good correspondence between observational flare signatures on high flaring rate stars and the Sun, which suggests a common physical origin. If the frequent and extreme flares that young solar-like stars and low-mass stars experience are accompanied by transient mass-loss in the form of coronal mass ejections, then the cumulative effect of this mass-loss could be large. We find that for young solar-like stars and active M dwarfs, the total mass lost due to transient magnetic eruptions could have significant impacts on disk evolution, and thus planet formation, and also exoplanet habitability.

Transient heat transfer for forced convection flow of helium gas

International Nuclear Information System (INIS)

Liu, Qiusheng; Fukuda, Katsuya; Sasaki, Kenji; Yamamoto, Manabu

1999-01-01

Transient heat transfer coefficients for forced convection flow of helium gas over a horizontal cylinder were measured using a forced convection test loop. The platinum heater with a diameter of 1.0 mm was heated by electric current with an exponential increase of Q 0 exp(t/τ). It was clarified that the heat transfer coefficient approaches the steady-state one for the period τ over 1 s, and it becomes higher for the period of τ shorter than 1 s. The transient heat transfer shows less dependent on the gas flowing velocity when the period becomes very shorter. Semi-empirical correlations for steady-state and transient heat transfer were developed based on the experimental data. (author)

Analysis of loss of offsite power transient using RELAP5/MOD1/NSC

International Nuclear Information System (INIS)

Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo

1986-01-01

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAP5/MOD1/NSC), based upon the sequence of events for the KNU1( Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximate the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV setting etc. are recognized to be important in the transient analyses on a bestestimate basis. (Author)

The concept of the sodium cooled small fast reactor 4S and the analyses of the loss of flow events

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Ueda, Nobuyuki; Koga, Tomonari; Matsumiya, Hisato

2007-01-01

loss of forced flow proceeds at full power. This sequence is referred to as an unprotected loss-of-flow accident (ULOF). The following became clear by the CERES analysis. In the case of PLOF, the temperature of the fuel and coolant and the cladding CDF value satisfy the criteria sufficiently. Sufficient heat removal by the RVACS and the IRACS was confirmed. The thermal capacity of the shielding surrounding the core was understood to have a major influence on the transient performance of the 4S-10M. In the case of ULOF, core power was reduced by the effect of the negative feedback, and the temperature of every part and the cladding CDF value satisfy the criteria sufficiently. The design validity of 4S in the view point of the event of the loss of flow was confirmed. And the thermal hydraulics characteristics became clear by analytical evaluations. (author)

Transient disturbance growth in flows over convex surfaces

Science.gov (United States)

Karp, Michael; Hack, M. J. Philipp

2017-11-01

Flows over curved surfaces occur in a wide range of applications including airfoils, compressor and turbine vanes as well as aerial, naval and ground vehicles. In most of these applications the surface has convex curvature, while concave surfaces are less common. Since monotonic boundary-layer flows over convex surfaces are exponentially stable, they have received considerably less attention than flows over concave walls which are destabilized by centrifugal forces. Non-modal mechanisms may nonetheless enable significant disturbance growth which can make the flow susceptible to secondary instabilities. A parametric investigation of the transient growth and secondary instability of flows over convex surfaces is performed. The specific conditions yielding the maximal transient growth and strongest instability are identified. The effect of wall-normal and spanwise inflection points on the instability process is discussed. Finally, the role and significance of additional parameters, such as the geometry and pressure gradient, is analyzed.

Transient two-dimensional flow in porous media

International Nuclear Information System (INIS)

Sharpe, L. Jr.

1979-01-01

The transient flow of an isothermal ideal gas from the cavity formed by an underground nuclear explosion is investigated. A two-dimensional finite element method is used in analyzing the gas flow. Numerical results of the pressure distribution are obtained for both the stemming column and the surrounding porous media

Development of a safety analysis code for molten salt reactors

International Nuclear Information System (INIS)

Zhang Dalin; Qiu Suizheng; Su Guanghui

2009-01-01

The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

CHF during flow rate, pressure and power transients in heated channels

International Nuclear Information System (INIS)

Celata, G.P.; Cumo, M.

1987-01-01

The behaviour of forced two-phase flows following inlet flow rate, pressure and power transients is presented here with reference to experiments performed with a R-12 loop. A circular duct, vertical test section (L = 2300 mm; D = 7.5 mm) instrumented with fluid (six) and wall (twelve) thermocouples has been employed. Transients have been carried out performing several values of flow decays (exponential decrease), depressurization rates (exponential decrease) and power inputs (step-wise increase). Experimental data have shown the complete inadequacy of steady-state critical heat flux correlations in predicting the onset of boiling crisis during fast transients. Data analysis for a better theoretical prediction of CHF occurrence during transient conditions has been accomplished, and design correlations for critical heat flux and time-to-crisis predictions have been proposed for the different types of transients

Surface chemistry of "unprotected" nanoparticles

DEFF Research Database (Denmark)

Schrader, Imke; Warneke, Jonas; Neumann, Sarah

2015-01-01

The preparation of colloidal nanoparticles in alkaline ethylene glycol is a powerful approach for the preparation of model catalysts and ligand-functionalized nanoparticles. For these systems the term "unprotected" nanoparticles has been established because no strongly binding stabilizers...... study. "Unprotected" Pt and Ru nanoparticles were characterized by NMR spectroscopy, which does not evidence the presence of any C-H containing species bound to the particle surface. Instead, the colloids were found to be covered by CO, as demonstrated by IR spectroscopy. However, analysis...

Characteristic behavior of bubbles and slugs in transient two-phase flow using image-processing method

International Nuclear Information System (INIS)

Goto, Shoji; Ishizaki, Yasuo; Ohashi, Hirotada; Akiyama, Mamoru

1995-01-01

Simulation of transient two-phase flow has been performed by solving transient hydrodynamic equations. However, constitution relations used in this simulation are primarily based on steady-state experimental results. Thus it is important to understand the transient behavior of bubbles and slugs, in particular, transient behavior of the void fraction, the interfacial area and the flow pattern, to confirm the applicability of the present simulation method and to advance two-phase flow simulation further. The present study deals with measurement of transient two-phase flow. We have measured local and instantaneous void fractions using imaging techniques, and compared the experimental data with simulation results. (author)

Modelling of an ULOF transient in a sodium fast reactor

International Nuclear Information System (INIS)

Droin, Jean-Baptiste

2016-01-01

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R and D program of CEA (French Commissariat a l'Energie Atomique et aux Energies Alternatives), safety in case of severe accidents is assessed.Such transients are usually simulated with mechanistic codes (such as SAS-SFR and SIMMER III). as a complement to these codes, which give reference accidental transient calculations, a new physico-statistical approach is currently followed by the CEA; its final objective being to derive the variability of the main results of interest for safety. This approach involves a fast-running description of extended accident sequences coupling physical models for the main phenomena to advanced statistical analysis techniques. It enables to perform a large number of simulations in a reasonable computational time and to describe all the possible bifurcations of the accident transient.In this context, this PhD work presents the physical tool (models and results assessment) dedicated to the initiation and primary phases of an Unprotected Loss Of Flow accident (i.e. until the end of sub-assemblies degradation and before large molten pools formation). The accident phenomenology during these phases is described and illustrated by numerous experimental evidences.It is underlined that the features of the new heterogeneous core concept (called CFV of the French ASTRID prototype) leads to different kinds of ULOF transients than those occurring in the previous past homogeneous cores (SuperPhenix, Phenix...). Indeed, its negative void effect drops the nuclear power when sodium heats-up and possibly boils. This enables three types of ULOF transients characterized by various core final states; the first two types leading to final coolable core states in natural circulation flow (the first one in single phase, the second one in stabilized two-phase flow) whereas the core undergoes a flow excursion followed by sub-assemblies degradation in the last type. In this study, a

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

Science.gov (United States)

Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

2017-11-01

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

CFD investigation of flow inversion in typical MTR research reactor undergoing thermal-hydraulic transients

International Nuclear Information System (INIS)

Salama, Amgad

2011-01-01

Highlights: → The 3D, CFD simulation of FLOFA accident in the generic IAEA 10 MW research reactor is carried out. → The different flow and heat transfer mechanisms involved in this process were elucidated. → The transition between these mechanisms during the course of FLOFA is discussed and investigated. → The interesting inversion process upon the transition from downward flow to upward flow is shown. → The temperature field and the friction coefficient during the whole transient process were shown. - Abstract: Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.

Simulation of transient fluid flow in mold region during steel continuous casting

International Nuclear Information System (INIS)

Liu, R; Thomas, B G; Sengupta, J

2012-01-01

A system of models has been developed to study transient flow during continuous casting and applied to simulate an event of multiple stopper-rod movements. It includes four sub-models to incorporate different aspects in this transient event. A three-dimensional (3-D) porous-flow model of the nozzle wall calculates the rate argon gas flow into the liquid steel, and the initial mean bubble size is estimated. Transient CFD models simulate multiphase flow of steel and gas bubbles in the Submerged Entry Nozzle (SEN) and mold and have been validated with experimental data from both nail dipping and Sub-meniscus Velocity Control (SVC) measurements. To obtain the transient inlet boundary conditions for the simulation, two semi-empirical models, a stopper-rod-position based model and a metal-level-based model, predict the liquid steel flow rate through the SEN based on recorded plant data. Finally the model system was applied to study the effects of stopper rod movements on SEN/mold flow patterns. Meniscus level fluctuations were calculated using a simple pressure method and compared well with plant measurements. Insights were gained from the simulation results to explain the cause of meniscus level fluctuations and the formation of sliver defects during stopper rod movements.

Comparison of the transient behavior of lead-based advanced critical and sub-critical reactors

International Nuclear Information System (INIS)

Wang Gang; Gu Zhixing; Wang Zhen; Jin Ming; Bai Yunqing

2014-01-01

A lead-based reactor developed by FDS Team is proposed in 2011 and designed to be 10 MW. It is a pool type reactor and the primary coolant is driven by natural circulation. The reactor has two operation modes, which are a lead-based critical fast reactor mode and a lead-based sub-critical reactor mode. The conceptual designs of the two modes are both completed by 2013. In this paper, four transient accidents were simulated for both the critical and sub-critical reactors above by NTC-2D code, which is developed by FDS Team for advanced reactor safety analysis. The four accidents were protected and unprotected loss of heat sink accidents (PLOHS and ULOHS), protected and unprotected transient overpower accidents (PTOP and UTOP). The simulation results of the two reactors were compared and analyzed. The results showed that during PLOHS and PTOP accidents for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the sub-critical reactor increased bigger than those of the critical one. For UTOP, the parameters above of the critical fast reactor were much bigger than those of the sub-critical one. The analysis results showed different safety advantages of the lead-based critical fast and sub-critical reactors during different transient accidents. (author)

Analysis of steady state and transient two-phase flows in downwardly inclined lines

International Nuclear Information System (INIS)

Crawford, T.J.

1983-01-01

A study of steady-state and transient two-phase flows in downwardly inclined lines is described. Steady-state flow patterns maps are presented using Freon-113 as the working fluid to provide new high density vapors. These flow maps with high density vapor serve to significantly extend the investigations of steady-state downward two-phase flow patterns. Physical models developed which successfully predicted the onset or location of various flow pattern transitions. A new simplified criterion that would be useful to designers and experimenters is offered for the onset of dispersed flow. A new empirical holdup correlation and a new bubble diameter/flow rate correlation are also proposed. Flow transients in vertical downward lines were studied to investigate the possible formation of intermediate or spurious flow patterns that would not be seen at steady-state conditions. Void fraction behavior during the transients was modeled by using the dynamic slip equation from the transient analysis code RETRAN. Physical models of interfacial area were developed and compared with models and data from literature. There was satisfactory agreement between the models of the present study and the literature models and data. The concentration parameter of the drift flux model was evaluated for vertical downward flow. These new values of the flow dependent parameter were different from those previously proposed in the literature for use in upward flows, and made the drift flux model suitable for use in upward or downward flow lines

PSB-VVER simulation of Kozloduy NPP 'loss of feed water transient'

Energy Technology Data Exchange (ETDEWEB)

Groudev, P.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg; Stefanova, A.E. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg; Gencheva, R.V. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: roseh@inrne.bas.bg; Pavlova, M.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlova@inrne.bas.bg

2005-04-01

This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions. RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient. The objective of the experiment 'loss of feed water', which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as 'integral system effects' and 'natural circulation'. For assessment of the RELAP5 capability to predict the 'Integral system effect' phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the 'Natural circulation' phenomenon the hot and cold leg temperatures behavior have been investigated. This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

International Nuclear Information System (INIS)

Mueller, C.J.

1986-01-01

The concept of inherent shutdown is emphasized in the approach to the design of innovative, small pool-type liquid-metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower events in evolving metal and oxide innovative designs

Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

International Nuclear Information System (INIS)

Mueller, C.J.

1986-01-01

The concept of ''inherent shutdown'' is emphasized in the approach to the design of innovative, small pool-type liquid metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram (ATWS) for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower (TOP) events in evolving metal and oxide innovative designs

Simulation of corrosion product activity in pressurized water reactors under flow rate transients

International Nuclear Information System (INIS)

Mirza, Anwar M.; Mirza, Nasir M.; Mir, Imran

1998-01-01

Simulation of coolant activation due to corrosion products and impurities in a typical pressurized water reactor has been done under flow rate transients. Employing time dependent production and losses of corrosion products in the primary coolant path an approach has been developed to calculate the coolant specific activity. Results for 24 Na, 56 Mn, 59 Fe, 60 Co and 99Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity is due to Mn-56. Parametric studies at full power for various ramp decreases in flow rate show initial decline in the activity and then a gradual rise to relatively higher saturation values. The minimum value and the time taken to reach the minima are strong functions of the slope of linear decrease in flow rate. In the second part flow rate coastdown was allowed to occur at different flow half-times. The reactor scram was initiated at 90% of the normal flow rate. The results show that the specific activity decreases and the rate of decrease depends on pump half time and the reactor scram conditions

Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (3) Identification of dominant factors in transition phase of unprotected events

International Nuclear Information System (INIS)

Tobita, Yoshiharu; Yamano, Hidemasa; Sato, Ikken

2009-01-01

The event progression of the transition phase in the unprotected loss of flow accident of the JSFR design concept was analyzed using the SIMMER-III code reflecting the knowledge obtained from the EAGLE experimental program. It was clarified through the parametric calculations that the fuel discharge behavior through the paths such as the inner duct of modified-FAIDUS and control-rod guide tube is playing a very important role. Effective fuel discharge through these paths prevents possibility of severe recriticality events. Important factors dominating the transition phase were identified through these parametric calculations. (author)

The PARET code and the analysis of the SPERT I transients

Energy Technology Data Exchange (ETDEWEB)

Woodruff, William L [Argonne National Laboratory, Argonne (United States)

1983-09-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.

The PARET code and the analysis of the SPERT I transients

International Nuclear Information System (INIS)

Woodruff, William L.

1983-01-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients

Response of EBR-II to a complete loss of primary forced flow during power operation

International Nuclear Information System (INIS)

Singer, R.M.; Gillette, J.L.; Mohr, D.; Tokar, J.V.; Sullivan, J.E.; Dean, E.M.

1980-01-01

Detailed measurements of the thermal, hydraulic, and neutronic response of EBR-II to a complete loss of primary forced flow followed by a PPS-activated scram are presented. The experimental results clearly indicate a smooth transition to natural convective flow with a quite modest incore temperature transient. The accompanying calculations using the NATDEMO code agree quite well with the measured temperatures and flow rates throughout the primary system. The only region of the plant where a significant discrepancy between the measurements and calculations occurred was in the IHX. The reasons for this result could not be definitively determined, but it is speculated that the one-dimensional assumptions used in the modeling may not be valid in the IHX during buoyancy driver flows

Transient gas flow through layered porous media

International Nuclear Information System (INIS)

Morrison, F.A. Jr.

1975-01-01

Low Reynolds number isothermal flow of an ideal gas through layered porous material was investigated analytically. Relations governing the transient flow in one dimension are obtained. An implicit, iterative, unconditionally stable finite difference scheme is developed for calculation of such flows. A computer code, SIROCCO, employing this technique has been written and implemented on the LLL computer system. A listing of the code is included. This code may be effectively applied to the evaluation of stemming plans for underground nuclear experiments. (U.S.)

A new transiently chaotic flow with ellipsoid equilibria

Science.gov (United States)

Panahi, Shirin; Aram, Zainab; Jafari, Sajad; Pham, Viet-Thanh; Volos, Christos; Rajagopal, Karthikeyan

2018-03-01

In this article, a simple autonomous transiently chaotic flow with cubic nonlinearities is proposed. This system represents some unusual features such as having a surface of equilibria. We shall describe some dynamical properties and behaviours of this system in terms of eigenvalue structures, bifurcation diagrams, time series, and phase portraits. Various behaviours of this system such as periodic and transiently chaotic dynamics can be shown by setting special parameters in proper values. Our system belongs to a newly introduced category of transiently chaotic systems: systems with hidden attractors. Transiently chaotic behaviour of our proposed system has been implemented and tested by the OrCAD-PSpise software. We have found a proper qualitative similarity between circuit and simulation results.

Study on numerical methods for transient flow induced by speed-changing impeller of fluid machinery

International Nuclear Information System (INIS)

Wu, Dazhuan; Chen, Tao; Wang, Leqin; Cheng, Wentao; Sun, Youbo

2013-01-01

In order to establish a reliable numerical method for solving the transient rotating flow induced by a speed-changing impeller, two numerical methods based on finite volume method (FVM) were presented and analyzed in this study. Two-dimensional numerical simulations of incompressible transient unsteady flow induced by an impeller during starting process were carried out respectively by using DM and DSR methods. The accuracy and adaptability of the two methods were evaluated by comprehensively comparing the calculation results. Moreover, an intensive study on the application of DSR method was conducted subsequently. The results showed that transient flow structure evolution and transient characteristics of the starting impeller are obviously affected by the starting process. The transient flow can be captured by both two methods, and the DSR method shows a higher computational efficiency. As an application example, the starting process of a mixed-flow pump was simulated by using DSR method. The calculation results were analyzed by comparing with the experiment data.

Numerical Investigation of Transient Flow in a Prototype Centrifugal Pump during Startup Period

Science.gov (United States)

Zhang, Yu-Liang; Zhu, Zu-Chao; Dou, Hua-Shu; Cui, Bao-Ling; Li, Yi; Zhou, Zhao-Zhong

2017-05-01

Transient performance of pumps during transient operating periods, such as startup and stopping, has drawn more and more attentions recently due to the growing engineering needs. During the startup period of a pump, the performance parameters such as the flow rate and head would vary significantly in a broad range. Therefore, it is very difficult to accurately specify the unsteady boundary conditions for a pump alone to solve the transient flow in the absence of experimental results. The closed-loop pipe system including a centrifugal pump is built to accomplish the self-coupling calculation. The three-dimensional unsteady incompressible viscous flow inside the passage of the pump during startup period is numerically simulated using the dynamic mesh method. Simulation results show that there are tiny fluctuations in the flow rate even under stable operating conditions and this can be attributed to influence of the rotor-stator interaction. At the very beginning of the startup, the rising speed of the flow rate is lower than that of the rotational speed. It is also found that it is not suitable to predict the transient performance of pumps using the calculation method of quasi-steady flow, especially at the earlier period of the startup.

FLATT - a computer programme for calculating flow and temperature transients in nuclear fuels

International Nuclear Information System (INIS)

Venkat Raj, V.; Koranne, S.M.

1976-01-01

FLATT is a computer code written in Fortran language for BESM-6 computer. The code calculates the flow transients in the coolant circuit of a nuclear reactor, caused by pump failure, and the consequent temperature transients in the fuel, clad, and the coolant. In addition any desired flow transient can be fed into the programme and the resulting temperature transients can be calculated. A case study is also presented. (author)

Analysis of three loss-of-flow accidents in the first wall cooling system of NET/ITER

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1993-05-01

This report presents the thermal-hydraulic analysis of three Loss-of-Flow Accidents (LOFAs) in the first wall cooling system of the Next European Torus (NET) design or the International Thermonuclear Experimental Reactor (ITER) design. The LOFAs considered result from a loss of the forced coolant flow caused by a loss of electrical power for the recirculation pump in the primary circuit. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analyses, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall. In the LOFA case without plasma shutdown, melting starts in the first wall about 150 s after accident initiation. In the LOFA case with delayed plasma shutdown, melting starts in the first wall when the plasma shutdown is initiated later than about 110 s after accident initiation. Melting does not occur in the first wall during a LOFA with prompt plasma scram. (orig.)

Uncertainty in simulated groundwater-quality trends in transient flow

Science.gov (United States)

Starn, J. Jeffrey; Bagtzoglou, Amvrossios; Robbins, Gary A.

2013-01-01

In numerical modeling of groundwater flow, the result of a given solution method is affected by the way in which transient flow conditions and geologic heterogeneity are simulated. An algorithm is demonstrated that simulates breakthrough curves at a pumping well by convolution-based particle tracking in a transient flow field for several synthetic basin-scale aquifers. In comparison to grid-based (Eulerian) methods, the particle (Lagrangian) method is better able to capture multimodal breakthrough caused by changes in pumping at the well, although the particle method may be apparently nonlinear because of the discrete nature of particle arrival times. Trial-and-error choice of number of particles and release times can perhaps overcome the apparent nonlinearity. Heterogeneous aquifer properties tend to smooth the effects of transient pumping, making it difficult to separate their effects in parameter estimation. Porosity, a new parameter added for advective transport, can be accurately estimated using both grid-based and particle-based methods, but predictions can be highly uncertain, even in the simple, nonreactive case.

Transient two-phase performance of LOFT reactor coolant pumps

International Nuclear Information System (INIS)

Chen, T.H.; Modro, S.M.

1983-01-01

Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

International Nuclear Information System (INIS)

Su'ud, Zaki

2008-01-01

In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

Transient simulation of hydropower station with consideration of three-dimensional unsteady flow in turbine

International Nuclear Information System (INIS)

Huang, W D; Fan, H G; Chen, N X

2012-01-01

To study the interaction between the transient flow in pipe and the unsteady turbulent flow in turbine, a coupled model of the transient flow in the pipe and three-dimensional unsteady flow in the turbine is developed based on the method of characteristics and the fluid governing equation in the accelerated rotational relative coordinate. The load-rejection process under the closing of guide vanes of the hydraulic power plant is simulated by the coupled method, the traditional transient simulation method and traditional three-dimensional unsteady flow calculation method respectively and the results are compared. The pressure, unit flux and rotation speed calculated by three methods show a similar change trend. However, because the elastic water hammer in the pipe and the pressure fluctuation in the turbine have been considered in the coupled method, the increase of pressure at spiral inlet is higher and the pressure fluctuation in turbine is stronger.

Transient simulation of hydropower station with consideration of three-dimensional unsteady flow in turbine

Science.gov (United States)

Huang, W. D.; Fan, H. G.; Chen, N. X.

2012-11-01

To study the interaction between the transient flow in pipe and the unsteady turbulent flow in turbine, a coupled model of the transient flow in the pipe and three-dimensional unsteady flow in the turbine is developed based on the method of characteristics and the fluid governing equation in the accelerated rotational relative coordinate. The load-rejection process under the closing of guide vanes of the hydraulic power plant is simulated by the coupled method, the traditional transient simulation method and traditional three-dimensional unsteady flow calculation method respectively and the results are compared. The pressure, unit flux and rotation speed calculated by three methods show a similar change trend. However, because the elastic water hammer in the pipe and the pressure fluctuation in the turbine have been considered in the coupled method, the increase of pressure at spiral inlet is higher and the pressure fluctuation in turbine is stronger.

ITER safety studies: The effect of two simultaneous perturbations during a loss of plasma control transient

International Nuclear Information System (INIS)

Rivas, J.C.; Dies, J.

2014-01-01

Highlights: •We have re-examined the methodology employed in the analysis of the “Loss of plasma transients in ITER” safety reference events. •We show the possible transient effects of a combined malfunction in external heating system and change in plasma confinement. •We show the possible transient effects of a combined malfunction in fuelling system and change in plasma confinement. •We have shown that new steady-states can be achieved that are potentially dangerous for the wall integrity. -- Abstract: The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. Conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer are used to determine the effects of such transients on wall integrity from a thermal point of view. In this contribution, progress in a “two simultaneous perturbations over plasma” approach to the analysis of the loss of plasma control transients in ITER is presented. The effect of variation in confinement time is now considered, and the consequences of this variation are shown over a n–T diagram. The study has been done with the aid of AINA 3.0 code. This code implements the same 0D plasma-1D wall scheme used in previous LOPC studies. The rationale of this study is that, once the occurrence of a loss of plasma transient has been assumed, and due to the uncertainties in plasma physics, it does not seem so unlikely to assume the possibility of finding a new confinement mode during the transient. The cases selected are intended to answer to the question “what would happen if an unexpected change in plasma confinement conditions takes place during a loss of plasma control transient due to a simultaneous malfunction of heating, or fuelling systems?” Even taking into account the simple models used and the uncertainties in plasma physics and design data, the

Influence of Base Oil Polarity on the Transient Shear Flow of Biodegradable Lubricating Greases

Directory of Open Access Journals (Sweden)

Martin Fiedler

2015-09-01

Full Text Available The scope of this study is to elucidate the physical mechanisms influencing the transient flow behavior of lubricating greases based on biogenic oleochemicals from a polarity point of view. This includes the mutually interacting influence of base oil polarity and thickening agents on the rheologically-measured mechanical structural degradation in transient shear flow. Due to the high temperature dependence of Keesom forces in the background of polar-active bond mechanisms, the analysis of the transient flow response as a function of temperature allows to attribute the observed influences to differences in base oil polarity. In general, clay-thickened greases show a greater tendency to be rheologically influenced by base oil polarities than soap-thickened lubricating greases.

Tackling complex turbulent flows with transient RANS

NARCIS (Netherlands)

Kenjeres, S.; Hanjalic, K.

2009-01-01

This article reviews some recent applications of the transient-Reynoldsaveraged Navier–Stokes (T-RANS) approach in simulating complex turbulent flows dominated by externally imposed body forces, primarily by thermal buoyancy and the Lorentz force. The T-RANS aims at numerical resolving unsteady

Transient flow characteristics of nuclear reactor coolant pump in recessive cavitation transition process

International Nuclear Information System (INIS)

Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun

2013-01-01

The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)

Experimental data report for transient flow calibration facility tests IIB101, IIB102 and IIB201

International Nuclear Information System (INIS)

Wambach, J.L.

1980-01-01

Thermal-hydraulic response data are presented for the transient performance tests of a pitot tube rake (IIB201) and a modular drag disc-turbine transducer (DTT) rake (IIB101, IIB102). The tests were conducted in a system which provided full scale simulation of the pressure vessel and broken loop hot leg piping of the Loss of Fluid Test Facility (LOFT). A load cell system was used to provide a reference mass flow rate measurement

A point implicit time integration technique for slow transient flow problems

Energy Technology Data Exchange (ETDEWEB)

Kadioglu, Samet Y., E-mail: kadioglu@yildiz.edu.tr [Department of Mathematical Engineering, Yildiz Technical University, 34210 Davutpasa-Esenler, Istanbul (Turkey); Berry, Ray A., E-mail: ray.berry@inl.gov [Idaho National Laboratory, P.O. Box 1625, MS 3840, Idaho Falls, ID 83415 (United States); Martineau, Richard C. [Idaho National Laboratory, P.O. Box 1625, MS 3840, Idaho Falls, ID 83415 (United States)

2015-05-15

Highlights: • This new method does not require implicit iteration; instead it time advances the solutions in a similar spirit to explicit methods. • It is unconditionally stable, as a fully implicit method would be. • It exhibits the simplicity of implementation of an explicit method. • It is specifically designed for slow transient flow problems of long duration such as can occur inside nuclear reactor coolant systems. • Our findings indicate the new method can integrate slow transient problems very efficiently; and its implementation is very robust. - Abstract: We introduce a point implicit time integration technique for slow transient flow problems. The method treats the solution variables of interest (that can be located at cell centers, cell edges, or cell nodes) implicitly and the rest of the information related to same or other variables are handled explicitly. The method does not require implicit iteration; instead it time advances the solutions in a similar spirit to explicit methods, except it involves a few additional function(s) evaluation steps. Moreover, the method is unconditionally stable, as a fully implicit method would be. This new approach exhibits the simplicity of implementation of explicit methods and the stability of implicit methods. It is specifically designed for slow transient flow problems of long duration wherein one would like to perform time integrations with very large time steps. Because the method can be time inaccurate for fast transient problems, particularly with larger time steps, an appropriate solution strategy for a problem that evolves from a fast to a slow transient would be to integrate the fast transient with an explicit or semi-implicit technique and then switch to this point implicit method as soon as the time variation slows sufficiently. We have solved several test problems that result from scalar or systems of flow equations. Our findings indicate the new method can integrate slow transient problems very

A point implicit time integration technique for slow transient flow problems

International Nuclear Information System (INIS)

Kadioglu, Samet Y.; Berry, Ray A.; Martineau, Richard C.

2015-01-01

Highlights: • This new method does not require implicit iteration; instead it time advances the solutions in a similar spirit to explicit methods. • It is unconditionally stable, as a fully implicit method would be. • It exhibits the simplicity of implementation of an explicit method. • It is specifically designed for slow transient flow problems of long duration such as can occur inside nuclear reactor coolant systems. • Our findings indicate the new method can integrate slow transient problems very efficiently; and its implementation is very robust. - Abstract: We introduce a point implicit time integration technique for slow transient flow problems. The method treats the solution variables of interest (that can be located at cell centers, cell edges, or cell nodes) implicitly and the rest of the information related to same or other variables are handled explicitly. The method does not require implicit iteration; instead it time advances the solutions in a similar spirit to explicit methods, except it involves a few additional function(s) evaluation steps. Moreover, the method is unconditionally stable, as a fully implicit method would be. This new approach exhibits the simplicity of implementation of explicit methods and the stability of implicit methods. It is specifically designed for slow transient flow problems of long duration wherein one would like to perform time integrations with very large time steps. Because the method can be time inaccurate for fast transient problems, particularly with larger time steps, an appropriate solution strategy for a problem that evolves from a fast to a slow transient would be to integrate the fast transient with an explicit or semi-implicit technique and then switch to this point implicit method as soon as the time variation slows sufficiently. We have solved several test problems that result from scalar or systems of flow equations. Our findings indicate the new method can integrate slow transient problems very

Transient Hearing Loss in Adults Associated With Zika Virus Infection.

Science.gov (United States)

Vinhaes, Eriko S; Santos, Luciane A; Dias, Lislane; Andrade, Nilvano A; Bezerra, Victor H; de Carvalho, Anderson T; de Moraes, Laise; Henriques, Daniele F; Azar, Sasha R; Vasilakis, Nikos; Ko, Albert I; Andrade, Bruno B; Siqueira, Isadora C; Khouri, Ricardo; Boaventura, Viviane S

2017-03-01

In 2015, during the outbreak of Zika virus (ZIKV) in Brazil, we identified 3 cases of acute hearing loss after exanthematous illness. Serology yielded finding compatible with ZIKV as the cause of a confirmed (n = 1) and a probable (n = 2) flavivirus infection, indicating an association between ZIKV infection and transient hearing loss. © The Author 2016. Published by Oxford University Press for the Infectious Diseases Society of America.

Transient performance of flow in PWR reactor circuits

International Nuclear Information System (INIS)

Hirdes, V.R.T.R.; Carajilescov, P.

1988-12-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Reduced-order modellin for high-pressure transient flow of hydrogen-natural gas mixture

Science.gov (United States)

Agaie, Baba G.; Khan, Ilyas; Alshomrani, Ali Saleh; Alqahtani, Aisha M.

2017-05-01

In this paper the transient flow of hydrogen compressed-natural gas (HCNG) mixture which is also referred to as hydrogen-natural gas mixture in a pipeline is numerically computed using the reduced-order modelling technique. The study on transient conditions is important because the pipeline flows are normally in the unsteady state due to the sudden opening and closure of control valves, but most of the existing studies only analyse the flow in the steady-state conditions. The mathematical model consists in a set of non-linear conservation forms of partial differential equations. The objective of this paper is to improve the accuracy in the prediction of the HCNG transient flow parameters using the Reduced-Order Modelling (ROM). The ROM technique has been successfully used in single-gas and aerodynamic flow problems, the gas mixture has not been done using the ROM. The study is based on the velocity change created by the operation of the valves upstream and downstream the pipeline. Results on the flow characteristics, namely the pressure, density, celerity and mass flux are based on variations of the mixing ratio and valve reaction and actuation time; the ROM computational time cost advantage are also presented.

Transient burnout under rapid flow reduction condition

International Nuclear Information System (INIS)

Iwamura, Takamichi

1987-01-01

Burnout characteristics were experimentally studied using uniformly heated tube and annular test sections under rapid flow reduction conditions. Observations indicated that the onset of burnout under a flow reduction transient is caused by the dryout of a liquid film on the heated surface. The decrease in burnout mass velocity at the channel inlet with increasing flow reduction rate is attributed to the fact that the vapor flow rate continues to increase and sustain the liquid film flow after the inlet flow rate reaches the steady-state burnout flow rate. This is because the movement of the boiling boundary cannot keep up with the rapid reduction of inlet flow rate. A burnout model for the local condition could be applied to the burnout phenomena with the flow reduction under pressures of 0.5 ∼ 3.9 MPa and flow reduction rates of 0.6 ∼ 35 %/s. Based on this model, a method to predict the burnout time under a flow reduction condition was presented. The calculated burnout times agreed well with experimental results obtained by some investigators. (author)

Loss-of-flow test L5 on FFTF-type irradiated fuel

International Nuclear Information System (INIS)

Simms, R.; Gehl, S.M.; Lo, R.K.; Rothman, A.B.

1978-03-01

Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (Pu, U)O 2 fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The preirradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting after the loss of flow. The test simulation corresponds to a scenario for FTR in which fuel in high-power-structure subassemblies slump, resulting in a power excursion. The remaining subassemblies are subjected to this power burst. Test L5 addressed the fuel-motion behavior of the subassemblies in this latter category. Data from test-vehicle sensors, hodoscope, and post-mortem examinations were used to construct the sequence of events within the test zone. From these observations, the fuel underwent a predominantly dispersive event just after reaching a peak power six times nominal at, or after, scram. The fuel motion was apparently driven by the release of entrained fission-product gases, since fuel vapor pressure was deliberately kept below significant levels for the transient. The test remains show a wide range of microstructural evolution, depending on the extent of heat deposition along the active fuel column. Extensive fuel swelling was also observed as a result of the lack of the cladding restraint. The results of the thermal-hydraulic calculations with the SAS3A code agreed qualitatively with the postmortem results with respect to the extent of the melting and the dispersal of cladding and fuel. However, the calculated times of certain events did not agree with the observed times

Unified fluid flow model for pressure transient analysis in naturally fractured media

International Nuclear Information System (INIS)

Babak, Petro; Azaiez, Jalel

2015-01-01

Naturally fractured reservoirs present special challenges for flow modeling with regards to their internal geometrical structure. The shape and distribution of matrix porous blocks and the geometry of fractures play key roles in the formulation of transient interporosity flow models. Although these models have been formulated for several typical geometries of the fracture networks, they appeared to be very dissimilar for different shapes of matrix blocks, and their analysis presents many technical challenges. The aim of this paper is to derive and analyze a unified approach to transient interporosity flow models for slightly compressible fluids that can be used for any matrix geometry and fracture network. A unified fractional differential transient interporosity flow model is derived using asymptotic analysis for singularly perturbed problems with small parameters arising from the assumption of a much smaller permeability of the matrix blocks compared to that of the fractures. This methodology allowed us to unify existing transient interporosity flow models formulated for different shapes of matrix blocks including bounded matrix blocks, unbounded matrix cylinders with any orthogonal crossection, and matrix slabs. The model is formulated using a fractional order diffusion equation for fluid pressure that involves Caputo derivative of order 1/2 with respect to time. Analysis of the unified fractional derivative model revealed that the surface area-to-volume ratio is the key parameter in the description of the flow through naturally fractured media. Expressions of this parameter are presented for matrix blocks of the same geometrical shape as well as combinations of different shapes with constant and random sizes. Numerical comparisons between the predictions of the unified model and those obtained from existing transient interporosity ones for matrix blocks in the form of slabs, spheres and cylinders are presented for linear, radial and spherical flow types for

Transient well flow in vertically heterogeneous aquifers

Science.gov (United States)

Hemker, C. J.

1999-11-01

A solution for the general problem of computing well flow in vertically heterogeneous aquifers is found by an integration of both analytical and numerical techniques. The radial component of flow is treated analytically; the drawdown is a continuous function of the distance to the well. The finite-difference technique is used for the vertical flow component only. The aquifer is discretized in the vertical dimension and the heterogeneous aquifer is considered to be a layered (stratified) formation with a finite number of homogeneous sublayers, where each sublayer may have different properties. The transient part of the differential equation is solved with Stehfest's algorithm, a numerical inversion technique of the Laplace transform. The well is of constant discharge and penetrates one or more of the sublayers. The effect of wellbore storage on early drawdown data is taken into account. In this way drawdowns are found for a finite number of sublayers as a continuous function of radial distance to the well and of time since the pumping started. The model is verified by comparing results with published analytical and numerical solutions for well flow in homogeneous and heterogeneous, confined and unconfined aquifers. Instantaneous and delayed drainage of water from above the water table are considered, combined with the effects of partially penetrating and finite-diameter wells. The model is applied to demonstrate that the transient effects of wellbore storage in unconfined aquifers are less pronounced than previous numerical experiments suggest. Other applications of the presented solution technique are given for partially penetrating wells in heterogeneous formations, including a demonstration of the effect of decreasing specific storage values with depth in an otherwise homogeneous aquifer. The presented solution can be a powerful tool for the analysis of drawdown from pumping tests, because hydraulic properties of layered heterogeneous aquifer systems with

Mode and climatic factors effect on energy losses in transient heat modes of transmission lines

Science.gov (United States)

Bigun, A. Ya; Sidorov, O. A.; Osipov, D. S.; Girshin, S. S.; Goryunov, V. N.; Petrova, E. V.

2018-01-01

Electrical energy losses increase in modern grids. The losses are connected with an increase in consumption. Existing models of electric power losses estimation considering climatic factors do not allow estimating the cable temperature in real time. Considering weather and mode factors in real time allows to meet effectively and safely the consumer’s needs to minimize energy losses during transmission, to use electric power equipment effectively. These factors increase an interest in the evaluation of the dynamic thermal mode of overhead transmission lines conductors. The article discusses an approximate analytic solution of the heat balance equation in the transient operation mode of overhead lines based on the least squares method. The accuracy of the results obtained is comparable with the results of solving the heat balance equation of transient thermal mode with the Runge-Kutt method. The analysis of mode and climatic factors effect on the cable temperature in a dynamic thermal mode is presented. The calculation of the maximum permissible current for variation of weather conditions is made. The average electric energy losses during the transient process are calculated with the change of wind, air temperature and solar radiation. The parameters having the greatest effect on the transmission capacity are identified.

On the application of the method of turbulent flow to transient and quasi-stationary flow calculations in high temperature reactors with spherical fuel elements

International Nuclear Information System (INIS)

Hoefer, I.

1980-12-01

For the calculation of flows in high-temperature reactors and of their temperature behavior the equations of the method of turbulent flow in the primitive form are derived for inhomogeneous regions. This system of equations is appropriate for the investigation of transient and quasi-stationary phenomena in pebble beds. By modification of the flow function in parallel arranged reflector channels a parallel flow can be simulated. For simplification the flow in region with a smaller pressure loss is assumed to be a potential flow. For the numerical solution of the time-dependent convective parts of the system of equations a number of explicit and implicit difference methods are compared. If the method using UP-WIND differences is taken to be an interpolation method the introduction of an extension becomes possible, which together with preliminary integration of the fictional terms allows to apply larger time steps. The algebraic system of equations for numerical calculation of a steady flow field also is established by formation of UP-WIND differences for the convective terms. By mathematical verification of some examples the applicability of the mathematical model for flow problems in pebble beds with forced or natural convection is shown. (orig.) [de

Transient critical heat flux under flow coast-down in vertical annulus with non-uniform heat flux distribution

International Nuclear Information System (INIS)

Moon, S.K.; Chun, S.Y.; Choi, K.Y.; Yang, S.K.

2001-01-01

An experimental study on transient critical heat flux (CHF) under flow coast-down has been performed for water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady state CHF. The transient CHF experiments have been performed for three kinds of flow transient modes based on the coast-down data of the Kori 3/4 nuclear power plant reactor coolant pump. Most of the CHFs occurred in the annular-mist flow regime. Thus, it means that the possible CHF mechanism might be the liquid film dryout in the annular-mist flow regime. For flow transient mode with the smallest flow reduction rate, the time-to-CHF is the largest. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to-CHF becomes large as the heat flux decreases. Usually, the critical mass flux is large for slow flow reduction. There is a pressure effect on the ratio of the transient CHF data to steady state CHF data. Some conventional correlations show relatively better CHF prediction results for high system pressure, high quality and slow transient modes than for low system pressure, low quality and fast transient modes. (author)

Effect of pipe insulation losses on a loss-of-heat sink accident for an LMR

International Nuclear Information System (INIS)

Horak, W.C.; Guppy, J.G.; Wood, P.M.

1985-01-01

The efficacy of pipe radiation losses as a heat sink during LOHS in a loop-type LMR plant is investigated. The Super System Code (SSC), which was modified to include pipe radiation losses, was used to simulate such an LOHS in an LMR plant. In order to enhance these losses, the pipes were assumed to be insulated by rock wool, a material whose thermal conductivity increases with increasing temperature. A transient was simulated for a total of eight days, during which the coolant temperatures peaked well below saturation conditions and then declined steadily. The coolant flow rate in the loop remained positive throughout the transient

The integral fast reactor

International Nuclear Information System (INIS)

Till, C.E.

1987-01-01

On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

Energy Technology Data Exchange (ETDEWEB)

Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

Stationary liquid fuel fast reactor SLFFR — Part II: Safety analysis

Energy Technology Data Exchange (ETDEWEB)

Jing, T.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

2016-12-15

Highlights: • A multi-channel safety analysis code named MUSA is developed for SLFFR transient analyses. • MUSA is verified against the SYS4A/SASSYS-1 code by simulating the ULOF accident for the advanced burner test reactor. • It is shown that SLFFR has a passive shutdown capability for double-fault, beyond-design-basis accidents UTOP, ULOHS and ULOF. - Abstract: Safety characteristics have been evaluated for the stationary liquid fuel fast reactor (SLFFR) proposed for effective burning of hazardous TRU elements of used nuclear fuel. In order to model the geometrical configuration and reactivity feedback mechanisms unique to SLFFR, a multi-channel safety analysis code named MUSA was developed. MUSA solves the time-dependent coupled neutronics and thermal-fluidic problems. The thermal-fluidic behavior of the core is described by representing the core with one-dimensional parallel channels. The primary heat transport system is modeled by connecting compressible volumes by liquid segments. A point kinetics model with six delayed neutron groups is used to represent the fission power transients. The reactivity feedback is estimated by combining the temperature and density variations of liquid fuel, structural material and sodium coolant with the corresponding axial distributions of reactivity worth in each individual thermal-fluidic channel. Preliminary verification tests with a conventional solid fuel reactor agreed well with the reference solutions obtained with the SAS4A/SASSYS-1 code. Transient analyses of SLFFR were performed for unprotected transient over-power (UTOP), unprotected loss of heat sink (ULOHS) and unprotected loss of flow (ULOF) accidents. The results showed that the thermal expansion of liquid fuel provides sufficiently large negative feedback reactivity for passive shutdown of UTOP and ULOHS. The ULOF transient is also terminated passively with the negative reactivity introduced by the gas expansion modules installed at the core periphery

Transient analysis for lead-bismuth-cooled accelerator-driven system proposed by JAEA

International Nuclear Information System (INIS)

Sugawara, T.; Nishihara, K.; Tsujimoto, K.

2015-01-01

It is supposed that an Accelerator-driven System (ADS) is safer than conventional critical reactors since an ADS is driven by the external neutron source in the subcritical state. In this study, the transient analyses for the lead-bismuth cooled ADS proposed by JAEA were performed using the SIMMER-III and RELAP5/mod3.2 codes to investigate the possibility of core damage. In this research, 3 accidents: the protected loss of heat sink, the protected overcooling and the unprotected blockage accident were considered as typical ADS accidents. Through these calculations, it was confirmed that all calculation results, except for the protected loss of heat sink, fulfilled the no-damage criteria. In the protected loss of heat sink, the cladding tube temperature reached its melting temperature after 18-21 hours, although the calculation condition was very conservative. These results have led to requirements to design a safety system of the ADS to decrease the frequencies of accidents. (authors)

Temperature transient response measurement in flowing water

International Nuclear Information System (INIS)

Rainbird, J.C.

1980-01-01

A specially developed procedure is described for determining the thermal transient response of thermocouples and other temperature transducers when totally immersed in flowing water. The high velocity heat transfer conditions associated with this facility enable thermocouple response times to be predicted in other fluids. These predictions can be confirmed by electrical analogue experiments. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Transient phenomena in multiphase flow

International Nuclear Information System (INIS)

Afgan, N.H.

1988-01-01

This book is devoted to formulation of the two-phase system. Emphasis is given to classical instantaneous equations of mass momentum and energy for local conditions and respective averaging procedures and their relevance to the structure of transfer laws. In formulating an equation for a two-velocity continuum, two-phase dispersed flow, two-velocity and local inertial effects associated with contraction and expansion of the mixture have been considered. Particular attention is paid to the effects of interface topology and area concentration as well as the latter's dependence on interfacial transfer laws. Also covered are low bubble concentrations in basic nonuniform unsteady flow where interactions between bubbles are negligible but where the effects of bubbles must still be considered. Special emphasis has been given to the pairwise interaction of the bubble and respective hydrodynamic equations describing the motion of a pair of spherical bubbles through a liquid This book introduces turbulence phenomena in two-phase flow and related problems of phase distribution in two-phase flow. This includes an extensive survey of turbulence and phase distribution models in transient two-phase flow. It is shown that if the turbulent structure of the continuous phase of bubbly two-phase is either measured or can be predicted, then the observed lateral phase distribution can be determined by using an multidimensional two-fluid model in which all lateral forces are properly modeled

ATHENA simulations of divertor pump trip and loss of heat sink transients for the GSSR

Energy Technology Data Exchange (ETDEWEB)

Sjoeberg, A

2001-04-01

The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that may occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a trip of the main circulation pump in the divertor cooling loop as well as a loss of heat sink, both initiated at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for these two transients have been evaluated and summarized in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. The results from the analyses indicate that for the pump trip transient the margin against overheating of critical highly loaded parts of the divertor cassette is small but seems sufficient. In case of the loss of heat sink transient the conservative analysis reveals that the pressurizer safety valve will be opened for an extended period of time and the long term transient development indicates a risk of completely filling up the pressurizer vessel. Thus the margins against jeopardizing the integrity of the divertor cooling system with the current design are for this case small but can for a long term operation at associate conditions pose a problem.

Integrated, digital experiment transient control and safety protection of an in-pile test

International Nuclear Information System (INIS)

Thomas, R.W.; Whitacre, R.F.; Klingler, W.B.

1982-01-01

The Sodium Loop Safety Facility experimental program has demonstrated that in-pile loop fuel failure transient tests can be digitally controlled and protected with reliability and precision. This was done in four nuclear experiments conducted in the Engineering Test Reactor operated by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Loop sodium flow and reactor power transients can be programmed to sponsor requirements and verified prior to the test. Each controller has redundancy, which reduces the effect of single failures occurring during test transients. Feedback and reject criteria are included in the reactor power control. Timed sequencing integrates the initiation of the controllers, programmed safety set-points, and other experiment actions (e.g., planned scram). Off-line and on-line testing is included. Loss-of-flow, loss-of-piping-integrity, boiling-window, transient-overpower, and local fault tests have been successfully run using this system

Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

2012-07-01

Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

LOFT/L9-3, Loss of Fluid Test, Anticipated Transients with Multiple Failures

International Nuclear Information System (INIS)

1992-01-01

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: This was the third of the NRC L9 series of experiments on Anticipated Transients with Multiple Failures. Loss-of-feedwater effects were studied. The experiment was conducted on 7 April 1982

Contribution to the theoretical study of transient two-phase flows

International Nuclear Information System (INIS)

Achard, J.L.

1978-12-01

The work presented in this paper has been given rise from the existence of violent boiling phenomena of the coolant that have been revealed by reactor safety studies with water and sodium. The aim as to describe in a basic mammer, one of these phenomena called ''chugging'' or ''choucage''. The experimental part of this work concerns two original works concerning the temperature measurement at the wall; a device is proposed to evaluate the contact resistance and the thermal inertia of the thermocouple; from the measurements that have been obtained, the flux the wall transfers to the flow and the temperature of the internal wall surface are deduced. A statistical method is developed for dispersed two-phase flow study, to establish: 1) a mass transfer law, 2) a law of change of the flow configuration. The proposed model contains: 1) for the dispersed phase (vapor bubbles), the basic momentum transport equations; 2) for the continuous phase (liquid), the transport equations of the classical formulation. The statistical formulation introduces the interaction phenomenon between the phases before applying the operation of the average (homogenization method); it allows to introduce the coalescence phenomena of bubbles. Finally, structures of exchange laws for transient laminar flows are proposed: transient linear momentum exchange law; possible structures of heat exchange laws [fr

LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient

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 first OECD LOFT experiment was conducted on February 20, 1983. It was designed to evaluate the generic PWR system response during a complete loss-of-feedwater transient. The objective of the experiment was to investigate the performance of primary 'feed and bleed' using a 'bleed' from the PORV and 'feed' from the HPIS to provide decay heat removal and system pressure reduction while maintaining the primary coolant inventory. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

A Guide for Using the Transient Ground-Water Flow Model of the Death Valley Regional Ground-Water Flow System, Nevada and California

Energy Technology Data Exchange (ETDEWEB)

Joan B. Blainey; Claudia C. Faunt, and Mary C. Hill

2006-05-16

This report is a guide for executing numerical simulations with the transient ground-water flow model of the Death Valley regional ground-water flow system, Nevada and California using the U.S. Geological Survey modular finite-difference ground-water flow model, MODFLOW-2000. Model inputs, including observations of hydraulic head, discharge, and boundary flows, are summarized. Modification of the DVRFS transient ground-water model is discussed for two common uses of the Death Valley regional ground-water flow system model: predictive pumping scenarios that extend beyond the end of the model simulation period (1998), and model simulations with only steady-state conditions.

Associations with Unprotected Sexual Behavior Among HIV-Infected Drinkers in Western Kenya.

Science.gov (United States)

Papas, Rebecca K; Gakinya, Benson N; Mwaniki, Michael M; Wu, Xiaotian K; Lee, Hana; Martino, Steve; Klein, Debra A; Sidle, John E; Loxley, Michelle P; Keter, Alfred K; Baliddawa, Joyce B; Maisto, Stephen A

2018-05-16

Approximately 71% of HIV-infected individuals live in sub-Saharan Africa. Alcohol use increases unprotected sex, which can lead to HIV transmission. Little research examines risky sex among HIV-infected individuals in East Africa who are not sex workers. The study purpose was to examine associations with unprotected sex in a high-risk sample of 507 HIV-infected sexually active drinkers in western Kenya. They were enrolled in a trial to reduce alcohol use. Past-month baseline alcohol use and sexual behavior were assessed using the Timeline Followback. A zero-inflated negative binomial model examined associations with occurrence and frequency of unprotected sex. Results showed heavy drinking days were significantly associated with unprotected sex occurrence across gender, and with unprotected sex frequency among women. Among women, transactional sex, alcohol-related sexual expectations, condom use self-efficacy, drinking-and-protected-sex days and age were associated with unprotected sex occurrence while alcohol-related sexual expectations, depressive symptoms and condom use self-efficacy were associated with unprotected sex frequency. Among men, alcohol-related sexual expectations, condom use self-efficacy, and age were associated with unprotected sex occurrence, while drinking-and-protected-sex days were associated with unprotected sex occurrence and frequency. Findings suggest robust relationships between heavy drinking and unprotected sex. Further research is needed elucidating the temporal relationships between drinking and unprotected sex in this population.

Critical heat flux phenomena in flow boiling during step wise and ramp wise power transients

International Nuclear Information System (INIS)

Celata, G.P.; Cumo, M.; D'Annibale, F.; Farello, G.E.; Abou Said, S.

1987-01-01

The present paper deals with the results of an experimental investigation of the forced flow critical heat flux during power transients in a vertically heated channel. Experiments were carried out with a Refrigerant-12 1oop employing a circular test section which was electrically and uniformly heated. The power transients were performed with the step-wise and ramp-wise increase of the power to the test section. The test parameters included several values of the initial power (before the transient) and the final power (at the end of the transient) in the case of step-wise transients and the slope of the ramp in the case of ramp-wise transients. The pressure and specific mass flow rate, which were kept constant during the power transient,were varied from 1.2 to 2.7 MPa and 850 to 1500 Kg/sm 2 , respectively. Correlations of the experimental data for the time-to-crisis in terms of the independent parameters of the system are also proposed and verified for different values of pressure,mass flow rate, and inlet subcooling

Transient analysis of air-water two-phase flow in channels and bends

International Nuclear Information System (INIS)

Khan, H.J.; Ye, W.; Pertmer, G.A.

1992-01-01

The algorithm used in this paper is the Newton Block Gauss Seidel method, which has been applied to both simple and complex flow conditions in two-phase flow. This paper contains a description of difference techniques and an iterative solution algorithm that is used to solve the field and constitutive equations of the two-fluid model. In practice, this solution procedure has been proven to be stable and capable of generating solutions in problems where other schemes have failed. The method converges rapidly for reasonable error tolerances and is easily extended to three-dimensional geometries. Using air-water as the two-phase medium, transient flow behavior in several geometries of interest are shown. Flow through a vertical channel with flow obstruction, large U bends, and 90-deg bends are being demonstrated with variation of inlet void fraction and slip ratio. Significant changes in the velocity and void distribution profiles have been observed. Various regions of flow recirculation are obtained in the flow domain for each phase. The phasic velocity and void distributions are dominated by gravity-induced phase separation causing air to accumulate in the upper region. The influence of inlet slip ratio and interfacial momentum transfer on the transient flow profile has been demonstrated in detail

Development and calibration of instruments for measurements in transient two-phase flow

International Nuclear Information System (INIS)

Banerjee, S.; Heidrick, T.R.

1981-01-01

For validation and development of theoretical models for transient two-phase flow, it is necessary to measure local and cross-sectionally averaged thermalhydraulic parameters. Of these parameters, void fraction and mass velocity are the most difficult to measure. In this paper, we present our recent work on various techniques for determining these quantities. The possibility of determining flow regime by using fast neutron transmission is discussed. The development of a miniaturized electrical resistivity probe for measuring local void fraction is described, together with calibrations obtained by integrating the void fraction profile and comparing the cross-sectionally averaged void fraction with direct measurements using two quick closing valves. Results on the calibration of combinations of full-flow turbine meters, Pitot tube rakes and gamma densitometers for measuring cross-sectionally averaged mass velocity in steady steam-water flow are presented. The results are interpreted with a simple model using single-phase calibration factors for the Pitot tube rakes and turbine meters. Calibration experiments were also done in transient steam-water flows and interpretation of the results with the steady state models is also discussed

Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

International Nuclear Information System (INIS)

Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

1991-01-01

A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

Study on transient hydrodynamic performance and cavitation characteristic of high-speed mixed-flow pump

International Nuclear Information System (INIS)

Chen, T; Liu, Y L; Sun, Y B; Wang, L Q; Wu, D Z

2013-01-01

In order to analyse the hydrodynamic performance and cavitation characteristic of a high-speed mixed-flow pump during transient operations, experimental studies were carried out. The transient hydrodynamic performance and cavitation characteristics of the mixed-flow pump with guide vane during start-up operation processes were tested on the pump performance test-bed. Performance tests of the pump were carried out under various inlet pressures and speed-changing operations. The real-time instantaneous external characteristics such as rotational speed, hydraulic head, flow rate, suction pressure and discharge pressure of the pump were measured. Based on the experimental results, the effect of fluid acceleration on the hydrodynamic performances and cavitation characteristics of the mixed-flow pump were analysed and evaluated

A large deviations approach to the transient of the Erlang loss model

NARCIS (Netherlands)

Mandjes, M.R.H.; Ridder, Annemarie

2001-01-01

This paper deals with the transient behavior of the Erlang loss model. After scaling both arrival rate and number of trunks, an asymptotic analysis of the blocking probability is given. Apart from that, the most likely path to blocking is given. Compared to Shwartz and Weiss [Large Deviations for

Measurement of transient hydrodynamic characteristics of the reactor RA primary cooling system

International Nuclear Information System (INIS)

Jovic, L.; Majstorovic, D.; Zeljkovic, I.

1987-01-01

Experimental study of transient hydrodynamic characteristics of the research nuclear reactor RA by simultaneous measurements of fluid flow and pressure on several locations of the RA primary coolant system is done. Loss of electric power transient on the main circulation pumps is simulated. measurement methodology, data processing and results of measured data analysis are given. (author)

The IFR status and prospects

International Nuclear Information System (INIS)

Till, C.E.

1992-01-01

The integral fast reactor (IFR) concept consists of a metal-fueled, liquid-metal-cooled fast reactor with its associated pyrometallurgical recovery and recycle of plutonium and higher actinides. The feasibility and operational stability of a liquid-metal-cooled fast reactor has been amply demonstrated in many countries. A comprehensive understanding of the irradiation effects on metal fuels permits extremely high fuel burnup without failures. Interest in the IFR concept grew rapidly when (in 1986) the benign response of the IFR was demonstrated by deliberately putting Experimental Breeder Reactor II (EBR-II) through unprotected loss-of-flow and loss-of-heat-sink transients; these transients were comfortably controlled by the inherent feedback of the system

Analysis of short-term reactor cavity transient

International Nuclear Information System (INIS)

Cheng, T.C.; Fischer, S.R.

1981-01-01

Following the transient of a hypothetical loss-of-coolant accident (LOCA) in a nuclear reactor, peak pressures are reached within the first 0.03 s at different locations inside the reactor cavity. Due to the complicated multidimensional nature of the reactor cavity, the short-term analysis of the LOCA transient cannot be performed by using traditional containment codes, such as CONTEMPT. The advanced containment code, BEACON/MOD3, developed at the Idaho National Engineering Laboratory (INEL), can be adapted for such analysis. This code provides Eulerian, one and two-dimensional, nonhomogeneous, nonequilibrium flow modeling as well as lumped parameter, homogeneous, equilibrium flow modeling for the solution of two-component, two-phase flow problems. The purpose of this paper is to demonstrate the capability of the BEACON code to analyze complex containment geometry such as a reactor cavity

Mountain scale modeling of transient, coupled gas flow, heat transfer and carbon-14 migration

International Nuclear Information System (INIS)

Lu, Ning; Ross, B.

1993-01-01

We simulate mountain-scale coupled heat transfer and gas flow at Yucca Mountain. A coupled rock-gas flow and heat transfer model, TGIF2, is used to simulate mountain-scale two-dimensional transient heat transfer and gas flow. The model is first verified against an analytical solution for the problem of an infinite horizontal layer of fluid heated from below. Our numerical results match very well with the analytical solution. Then, we obtain transient temperature and gas flow distributions inside the mountain. These distributions are used by a transient semianalytical particle tracker to obtain carbon-14 travel times for particles starting at different locations within the repository. Assuming that the repository is filled with 30-year-old waste at an initial areal power density of 57 kw/acre, we find that repository temperatures remain above 60 degrees C for more than 10,000 years. Carbon-14 travel times to the surface are mostly less than 1000 years, for particles starting at any time within the first 10,000 years

The ionospheric signature of transient dayside reconnection and the associated pulsed convection return flow

Directory of Open Access Journals (Sweden)

S. E. Milan

Full Text Available Three SuperDARN coherent HF radars are employed to investigate the excitation of convection in the dayside high-latitude ionosphere in response to transient reconnection occurring in the cusp region. This study demonstrates the existence of transient antisunward-propagating backscatter features at the expected location of the ionospheric footprint of the cusp region, which have a repetition rate near 10 min. These are interpreted as the ionospheric signature of flux transfer events. Moreover, transient sunward-propagating regions of backscatter are observed in the convection return flow regions of both the pre- and post-noon sectors. These patches are observed to propagate towards the noon sector from at least as far around the auroral zone as 07 MLT in the pre-noon sector and 17 MLT in the post-noon sector, travelling with a velocity of approximately 1.5 to 2 km s^-1. These return flow patches have a repetition rate similar to that of the transient features observed at local noon. While providing supporting evidence for the impulsive nature of convection flow, the observation of sunward-propagating features in the return flow region is not consistent with current conceptual models of the excitation of convection.

Key words. Ionosphere (plasma convection · Magnetospheric physics (magnetopause · cusp · and boundary layers; magnetosphere-ionosphere interactions

Performance evaluation of a drag-disc turbine transducer and three-beam gamma densitometer under transient two-phase flow conditions

International Nuclear Information System (INIS)

Nalezny, C.L.; Chen, L.L.; Solbrig, C.W.

1979-01-01

One of the primary variables measured in the Loss-of-Fluid Test (LOFT) Program is mass flow rate. LOFT uses drag-disc turbine tranducers (DTT) and a three-beam gamma densitometer to measure parameters from which mass flow may be computed. The transducer combination was performance tested under transient conditions in the blowdown loop at the LOFT Test Support Facility (LTSF). The performance tests consisted of three partial blowdowns of different durations starting from the same initial conditions. The reference mean mass flow rate was determined by measuring the amount of water required to reestablish initial conditions after each partial blowdown. The average mass flow rates computed from the output of the drag disc, turbine, and gamma densitometer were compared to the reference mean mass flow rates over three blowdown intervals. The tests indicated that the equal phase velocity mass measurement model provided excellent results through the use of the turbine and densitometer, and drag disc and densitometer when the phase velocities were nearly equal

Transient performance of flow in circuits of PWR type reactors

International Nuclear Information System (INIS)

Hirdes, V.R.; Carajilescov, P.

1988-09-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Fractional governing equations of transient groundwater flow in confined aquifers with multi-fractional dimensions in fractional time

Directory of Open Access Journals (Sweden)

M. L. Kavvas

2017-10-01

Full Text Available Using fractional calculus, a dimensionally consistent governing equation of transient, saturated groundwater flow in fractional time in a multi-fractional confined aquifer is developed. First, a dimensionally consistent continuity equation for transient saturated groundwater flow in fractional time and in a multi-fractional, multidimensional confined aquifer is developed. For the equation of water flux within a multi-fractional multidimensional confined aquifer, a dimensionally consistent equation is also developed. The governing equation of transient saturated groundwater flow in a multi-fractional, multidimensional confined aquifer in fractional time is then obtained by combining the fractional continuity and water flux equations. To illustrate the capability of the proposed governing equation of groundwater flow in a confined aquifer, a numerical application of the fractional governing equation to a confined aquifer groundwater flow problem was also performed.

HYTRAN: hydraulic transient code for investigating channel flow stability

International Nuclear Information System (INIS)

Kao, H.S.; Cardwell, W.R.; Morgan, C.D.

1976-01-01

HYTRAN is an analytical program used to investigate the possibility of hydraulic oscillations occurring in a reactor flow channel. The single channel studied is ordinarily the hot channel in the reactor core, which is parallel to other channels and is assumed to share a constant pressure drop with other channels. Since the channel of highest thermal state is studied, provision is made for two-phase flow that can cause a flow instability in the channel. HYTRAN uses the CHATA(1) program to establish a steady-state condition. A heat flux perturbation is then imposed on the channel, and the flow transient is calculated as a function of time

Safety studies: Review of loss of plasma control transients in ITER with AINA 3.0 code

International Nuclear Information System (INIS)

Rivas, J.C.; Dies, J.

2013-01-01

Highlights: ► We have examined the methodology employed in the analysis of the “Loss of plasma transients in ITER” safety reference events. ► We have developed a new methodology based on the study of the plasma operating window. ► We have concluded that the combined effect of different perturbations should be studied also to determine the most severe transients. -- Abstract: The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. In the past, conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer showed that a hypothetical scenario of first wall coolant tubes melting and subsequent entering of water in the vacuum vessel could not be totally excluded. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering Laboratory (FEEL) in Barcelona. It uses a 0D–1D architecture, similar to that used for previous analyses of ITER loss of plasma control events. The results of this study show the simultaneous effect of two perturbations (overfuelling and overheating) over a plasma transient, and compare it with the isolated effects of each perturbation. It is shown that the combined effect can be more severe, and a method is outlined to locate the most dangerous transients over a nT diagram

SCDAP/RELAP5 modeling of fluid heat transfer and flow losses through porous debris in a light water reactor

International Nuclear Information System (INIS)

Harvego, E. A.; Siefken, L. J.

2000-01-01

The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate fluid heat transfer and flow losses through porous debris that has accumulated in the vessel lower head and core regions during the latter stages of a severe accident. The implementation of heat transfer and flow loss correlations into the code is discussed, and calculations performed to assess the validity of the modeling approach are described. The different modes of heat transfer in porous debris include: (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, (5) film boiling, and (6) transition from film boiling to convection to vapor. The correlations for flow losses in porous debris include frictional and form losses. The correlations for flow losses were integrated into the momentum equations in the RELAP5 part of the code. Since RELAP5 is a very general non-homogeneous non-equilibrium thermal-hydraulics code, the resulting modeling methodology is applicable to a wide range of debris thermal-hydraulic conditions. Assessment of the SCDAP/RELAP5 debris bed thermal-hydraulic models included comparisons with experimental measurements and other models available in the open literature. The assessment calculations, described in the paper, showed that SCDAP/RELAP5 is capable of calculating the heat transfer and flow losses occurring in porous debris regions that may develop in a light water reactor during a severe accident

The role of natural circulation in the FFTF [Fast Flux Test Facility] passive safety tests

International Nuclear Information System (INIS)

Stover, R.L.; Padilla, A.; Burke, T.M.; Knecht, W.L.

1987-03-01

A series of tests were completed at the Fast Flux Test Facility to demonstrate the passive safety characteristics of liquid metal reactors with natural circulation flow. The first test consisted of transition from forced to natural circulation flow at an initial decay power of 0.3%. The second test represented an unprotected loss-of-flow transient to natural circulation from 50% power with the control rods prevented from scramming into the core. The third test was a steady-state, natural circulation condition with core fission powers up ato about 2.3%. Core sodium data and results of single and multi-channel computer models confirmed the reliability and effectiveness of natural circulation flow for liquid metal reactor safety

Analysis and computer simulation for transient flow in complex system of liquid piping

International Nuclear Information System (INIS)

Mitry, A.M.

1985-01-01

This paper is concerned with unsteady state analysis and development of a digital computer program, FLUTRAN, that performs a simulation of transient flow behavior in a complex system of liquid piping. The program calculates pressure and flow transients in the liquid filled piping system. The analytical model is based on the method of characteristics solution to the fluid hammer continuity and momentum equations. The equations are subject to wide variety of boundary conditions to take into account the effect of hydraulic devices. Water column separation is treated as a boundary condition with known head. Experimental tests are presented that exhibit transients induced by pump failure and valve closure in the McGuire Nuclear Station Low Level Intake Cooling Water System. Numerical simulation is conducted to compare theory with test data. Analytical and test data are shown to be in good agreement and provide validation of the model

The integral fast reactor

International Nuclear Information System (INIS)

Till, C.E.

1987-01-01

On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

Multidimensional Discriminative Factors for Unprotected Sex Among Adolescents in Southern Taiwan

Directory of Open Access Journals (Sweden)

Cheng-Fang Yen

2009-04-01

Full Text Available Establishing the discriminative factors for unprotected sex among adolescents is essential for early identification of at-risk teens and for the prevention of unplanned pregnancy and sexually transmitted diseases. The aim of this study was to examine the discriminative effects of demographic, individual, family, peers, and school life factors on unprotected sex in a large-scale, representative adolescent population in Southern Taiwan. A total of 9,736 adolescent students were recruited into this study and completed the questionnaires. The multidimensional discriminative factors for unprotected sex were examined using χ2 automatic interaction detection analysis and logistic regression models. The results of the χ2 automatic interaction detection analysis revealed that having friends, using illicit drugs, being of an older age, suspension from school, and low family monitoring had discriminative effects on unprotected sex in adolescents. The logistic regression analysis further confirmed the discriminative effect of these factors. Because of the adverse effects of unprotected sex in adolescents, we suggest that parents and health professionals should pay attention to adolescents with the discriminative factors for unprotected sex identified in this study.

Scaling of two-phase flow transients using reduced pressure system and simulant fluid

International Nuclear Information System (INIS)

Kocamustafaogullari, G.; Ishii, M.

1987-01-01

Scaling criteria for a natural circulation loop under single-phase flow conditions are derived. Based on these criteria, practical applications for designing a scaled-down model are considered. Particular emphasis is placed on scaling a test model at reduced pressure levels compared to a prototype and on fluid-to-fluid scaling. The large number of similarty groups which are to be matched between modell and prototype makes the design of a scale model a challenging tasks. The present study demonstrates a new approach to this clasical problen using two-phase flow scaling parameters. It indicates that a real time scaling is not a practical solution and a scaled-down model should have an accelerated (shortened) time scale. An important result is the proposed new scaling methodology for simulating pressure transients. It is obtained by considerung the changes of the fluid property groups which appear within the two-phase similarity parameters and the single-phase to two-phase flow transition prameters. Sample calculations are performed for modeling two-phase flow transients of a high pressure water system by a low-pressure water system or a Freon system. It is shown that modeling is possible for both cases for simulation pressure transients. However, simulation of phase change transitions is not possible by a reduced pressure water system without distortion in either power or time. (orig.)

Transient three-phase three-component flow. Pt. 3

International Nuclear Information System (INIS)

Kolev, N.I.

1986-05-01

A mathematical model of a transient three-dimensional three-phase three-component flow described by three-velocity fields in porous body is presented. A combination of separated mass and energy equations together with mixture momentum equations for the flow is used. The mixture equations are used in diffusion form with the assumption that the diffusion velocity can be calculated from empirical correlations. An analytical coupling between the governing equations is developed for calculation of the pressure field. The system is discretized semiimplicitly in 3D-cylindrical space and different solution methods for the algebraic problem are presented. Finally, numerical examples and comparisons with experimental data demonstrate that the method presented is a powerful tool for numerical multiphase flow simulation. (orig.) [de

Transient well flow in layered aquifer systems: the uniform well-face drawdown solution.

NARCIS (Netherlands)

Hemker, C.J.

1999-01-01

Previously a hybrid analytical-numerical solution for the general problem of computing transient well flow in vertically heterogeneous aquifers was proposed by the author. The radial component of flow was treated analytically, while the finite-difference technique was used for the vertical flow

Experimental data report for transient flow calibration facility tests IIIA101, IIIA102, IIIA201, and IIIA202

International Nuclear Information System (INIS)

Wambach, J.L.

1980-01-01

Thermal-hydraulic response data are presented for the transient performance tests of an ECC pitot tube rake (IIIA201, IIIA202) and both an ECC pitot tube rake and modular drag disc-turbine transducer (DTT) rake (IIIA101, IIIA102). The tests were conducted in a system which provided full scale simulation of the pressure vessel and intact loop cold leg piping of the Loss of Fluid Test Facility (LOFT). A load cell system was used to provide a reference mass flow rate measurement

Comparison of SAS3A and MELT-III predictions for a transient overpower hypothetical accident

International Nuclear Information System (INIS)

Wilburn, N.P.

1976-01-01

A comparison is made of the predictions of the two major codes SAS3A and MELT-III for the hypothetical unprotected transient overpower accident in the FFTF. The predictions of temperatures, fuel restructuring, fuel melting, reactivity feedbacks, and core power are compared

Death Valley regional groundwater flow system, Nevada and California-Hydrogeologic framework and transient groundwater flow model

Science.gov (United States)

Belcher, Wayne R.; Sweetkind, Donald S.

2010-01-01

A numerical three-dimensional (3D) transient groundwater flow model of the Death Valley region was developed by the U.S. Geological Survey for the U.S. Department of Energy programs at the Nevada Test Site and at Yucca Mountain, Nevada. Decades of study of aspects of the groundwater flow system and previous less extensive groundwater flow models were incorporated and reevaluated together with new data to provide greater detail for the complex, digital model. A 3D digital hydrogeologic framework model (HFM) was developed from digital elevation models, geologic maps, borehole information, geologic and hydrogeologic cross sections, and other 3D models to represent the geometry of the hydrogeologic units (HGUs). Structural features, such as faults and fractures, that affect groundwater flow also were added. The HFM represents Precambrian and Paleozoic crystalline and sedimentary rocks, Mesozoic sedimentary rocks, Mesozoic to Cenozoic intrusive rocks, Cenozoic volcanic tuffs and lavas, and late Cenozoic sedimentary deposits of the Death Valley regional groundwater flow system (DVRFS) region in 27 HGUs. Information from a series of investigations was compiled to conceptualize and quantify hydrologic components of the groundwater flow system within the DVRFS model domain and to provide hydraulic-property and head-observation data used in the calibration of the transient-flow model. These studies reevaluated natural groundwater discharge occurring through evapotranspiration (ET) and spring flow; the history of groundwater pumping from 1913 through 1998; groundwater recharge simulated as net infiltration; model boundary inflows and outflows based on regional hydraulic gradients and water budgets of surrounding areas; hydraulic conductivity and its relation to depth; and water levels appropriate for regional simulation of prepumped and pumped conditions within the DVRFS model domain. Simulation results appropriate for the regional extent and scale of the model were provided

Large Eddy Simulation of Transient Flow, Solidification, and Particle Transport Processes in Continuous-Casting Mold

Science.gov (United States)

Liu, Zhongqiu; Li, Linmin; Li, Baokuan; Jiang, Maofa

2014-07-01

The current study developed a coupled computational model to simulate the transient fluid flow, solidification, and particle transport processes in a slab continuous-casting mold. Transient flow of molten steel in the mold is calculated using the large eddy simulation. An enthalpy-porosity approach is used for the analysis of solidification processes. The transport of bubble and non-metallic inclusion inside the liquid pool is calculated using the Lagrangian approach based on the transient flow field. A criterion of particle entrapment in the solidified shell is developed using the user-defined functions of FLUENT software (ANSYS, Inc., Canonsburg, PA). The predicted results of this model are compared with the measurements of the ultrasonic testing of the rolled steel plates and the water model experiments. The transient asymmetrical flow pattern inside the liquid pool exhibits quite satisfactory agreement with the corresponding measurements. The predicted complex instantaneous velocity field is composed of various small recirculation zones and multiple vortices. The transport of particles inside the liquid pool and the entrapment of particles in the solidified shell are not symmetric. The Magnus force can reduce the entrapment ratio of particles in the solidified shell, especially for smaller particles, but the effect is not obvious. The Marangoni force can play an important role in controlling the motion of particles, which increases the entrapment ratio of particles in the solidified shell obviously.

ROSA-III/971, BWR Rig of Safety Assessment LOCA, Loss of Offsite Power Transient

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: ROSA-III is a 1/124 scaled down test facility with electrically heated core designed to study the response of engineered safety features to loss-of-coolant accidents in in commercial BWR. It consists of the following, fully instrumented subsystems: (a) the pressure vessel with a core simulating four half-length fuel assemblies and control rod; (b) steam line and feed water line, which are independent open loops; (c) coolant recirculation system, which consists of two loops provided with a recirculation pump and two jet pumps in each loop; (d) emergency cooling system, including HPCS, LPCS, LPCI, and ADS. 2 - Description of test: Run 971 simulated a BWR LOSS of off-site power transient. The core scram was assumed to occur at 6 seconds after the transient initiated by the turbine trip. HPCS failure was assumed. After ADS started, the upper half of the core was uncovered by steam. The core was re-flooded by LPCS alone

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays

International Nuclear Information System (INIS)

Benedetti, R.L.; Lords, L.V.; Kiser, D.M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage

Longitudinal influences of friends and parents upon unprotected vaginal intercourse in adolescents.

Science.gov (United States)

Kim, Catherine; Gebremariam, Acham; Iwashyna, Theodore J; Dalton, Vanessa K; Lee, Joyce M

2011-02-01

Both friends and parents may influence occurrence of adolescent sexual intercourse, but these influences have not been studied together and prospectively. We conducted a longitudinal analysis of a nationally representative sample of adolescents aged 15-18 years (n=6649), the National Longitudinal Study of Adolescent Health (Add Health). Baseline in-home and school interviews were conducted during 1995 and follow-up interviews in 1996. The main outcome measure was self-reported unprotected vaginal intercourse. In models which adjusted for age, race, parental attitudes towards contraception and pregnancy, and adolescent sexual intercourse practices at baseline, having a friend who engaged in sexual intercourse at baseline, either unprotected (OR 2.2, 95% CI 1.6-3.2) or protected (OR 1.8, 95% CI 1.4-2.4), increased the odds of unprotected intercourse vs. never intercourse in the adolescent at follow-up (pParental attitudes were not associated with the outcome after consideration of the adolescent's attitudes and baseline sexual practices. Having a friend who engages in sexual intercourse, unprotected or protected, increases the risk of unprotected intercourse. Parental attitudes are less influential after consideration of adolescent baseline attitudes and sexual practices, suggesting that parental influences are strongest before 15 years of age. Our results suggest that early intervention among both parents and adolescents may decrease the risk of unprotected intercourse. Copyright © 2011 Elsevier Inc. All rights reserved.

Influence of Flow Velocity on Tsunami Loss Estimation

Directory of Open Access Journals (Sweden)

Jie Song

2017-11-01

Full Text Available Inundation depth is commonly used as an intensity measure in tsunami fragility analysis. However, inundation depth cannot be taken as the sole representation of tsunami impact on structures, especially when structural damage is caused by hydrodynamic and debris impact forces that are mainly determined by flow velocity. To reflect the influence of flow velocity in addition to inundation depth in tsunami risk assessment, a tsunami loss estimation method that adopts both inundation depth and flow velocity (i.e., bivariate intensity measures in evaluating tsunami damage is developed. To consider a wide range of possible tsunami inundation scenarios, Monte Carlo-based tsunami simulations are performed using stochastic earthquake slip distributions derived from a spectral synthesis method and probabilistic scaling relationships of earthquake source parameters. By focusing on Sendai (plain coast and Onagawa (ria coast in the Miyagi Prefecture of Japan in a case study, the stochastic tsunami loss is evaluated by total economic loss and its spatial distribution at different scales. The results indicate that tsunami loss prediction is highly sensitive to modelling resolution and inclusion of flow velocity for buildings located less than 1 km from the sea for Sendai and Onagawa of Miyagi Prefecture.

An analysis of transient flow in upland watersheds: interactions between structure and process

Science.gov (United States)

David Lawrence Brown

1995-01-01

The physical structure and hydrological processes of upland watersheds interact in response to forcing functions such as rainfall, leading to storm runoff generation and pore pressure evolution. Transient fluid flow through distinct flow paths such as the soil matrix, macropores, saprolite, and bedrock may be viewed as a consequence of such interactions. Field...

Transient flows occurring during the accelerated crucible rotation technique

International Nuclear Information System (INIS)

Horowitz, Atara; Horowitz, Yigal

1992-11-01

The transient flows occurring after a change in the angular velocity of the cylindrical container are described. The dependence of the transient (known as spin-up or spin-down time) on experimental parameters as kinematic viscosity, cylinder dimensions and the cylinder's initial and final angular velocities are elucidates by a review of the literature. It is emphasized that with large Rossby numbers the spin-up time is longer and the amount of fluid mixing is greater than small and moderate Rossby numbers. It is also elucidated that most crystal growth crucibles cannot be considered as infinitely-long cylinders for the evaluation of the fluid dynamics (authors)

Prandtl number variation on transient forced convection flow in a ...

African Journals Online (AJOL)

user

2Manufacturing Engineering Department, The Public Authority for Applied Education and ... A transient numerical study is conducted to investigate the transport .... The model describes a valve where it is possible to direct the flow into one.

Modeling transient thermal hydraulic behavior of a thermionic fuel element for nuclear space reactors

International Nuclear Information System (INIS)

Al-Kheliewi, A.S.; Klein, A.C.

1994-01-01

A transient code (TFETC) for determining the temperature distribution throughout the radial and axial positions of a thermionic fuel element (TFE) during changes in operating conditions has been successfully developed and tested. A fully implicit method is used to solve the system of equations for temperatures at each time step. Presently, TFETC has the ability to handle the following transients: startup, loss of flow accidents, and shutdown. The code has been applied to the startup of the ATI single cell configuration which appears to start up and shut down in an orderly and reasonable fashion. No unexpected transient features were observed. The TFE also appears to function robustly under loss of flow accident conditions. It appears hat sufficient time is available to shut the reactor down safely without melting point the fuel. The model shows that during a complete loss of flow accident (without shutdown) the coolant reaches its boiling point in approximately 35 seconds. The fuel may exceed its melting point after this time as the NaK coolant will boil if the reactor is not shut down. For 1/2, 1/3, and 1/4 pump failures, the fuel temperatures never exceed the fuel melting point even if the reactor is not shut down

A Case of Recurrent Transient Monocular Visual Loss after Receiving Sildenafil

Directory of Open Access Journals (Sweden)

Asaad Ghanem Ghanem

2011-01-01

Full Text Available A 53-year-old man was attended to the Clinic Ophthalmic Center, Mansoura University, Egypt, with recurrent transient monocular visual loss after receiving sildenafil citrate (Viagra for erectile dysfunction. Examination for possible risk factors revealed mild hypercholesterolemia. Family history showed that his father had suffered from bilateral nonarteritic anterior ischemic optic neuropathy (NAION. Physicians might look for arteriosclerotic risk factors and family history of NAION among predisposing risk factors before prescribing sildenafil erectile dysfunction drugs.

Transient Characteristics of a Fluidic Device for Circulatory Jet Flow.

Science.gov (United States)

Phan, Hoa Thanh; Dinh, Thien Xuan; Bui, Phong Nhu; Dau, Van Thanh

2018-03-13

In this paper, we report on the design, simulation, and experimental analysis of a miniaturized device that can generate multiple circulated jet flows. The device is actuated by a lead zirconate titanate (PZT) diaphragm. The flows in the device were studied using three-dimensional transient numerical simulation with the programmable open source OpenFOAM and was comparable to the experimental result. Each flow is verified by two hotwires mounted at two positions inside each consisting chamber. The experiment confirmed that the flow was successfully created, and it demonstrated good agreement with the simulation. In addition, a prospective application of the device as an angular rate sensor is also demonstrated. The device is robust, is minimal in size, and can contribute to the development of multi-axis fluidic inertial sensors, fluidic amplifiers, gas mixing, coupling, and analysis.

Fractional governing equations of transient groundwater flow in confined aquifers with multi-fractional dimensions in fractional time

OpenAIRE

M. L. Kavvas; T. Tu; A. Ercan; J. Polsinelli

2017-01-01

Using fractional calculus, a dimensionally consistent governing equation of transient, saturated groundwater flow in fractional time in a multi-fractional confined aquifer is developed. First, a dimensionally consistent continuity equation for transient saturated groundwater flow in fractional time and in a multi-fractional, multidimensional confined aquifer is developed. For the equation of water flux within a multi-fractional multidimensional confined aquifer, a dimensionally...

Computer simulation of fuel behavior during loss-of-flow accidents in a gas-cooled fast reactor

International Nuclear Information System (INIS)

Wehner, T.R.

1980-01-01

The sequence of events in a loss-of-flow accident without reactor shutdown in a gas-cooled fast breeder reactor is strongly influenced by the manner in which the fuel deforms. In order to predict the mode of initial gross fuel deformation, welling, melting or cracking, a thermomechanical computer simulation program was developed. Methods and techniques used make the simulation an economical, efficient, and flexible engineering tool. An innovative application of the enthalpy model within a finite difference scheme is used to caculate temperatures in the fuel rod. The method of successive elastic solutions is used to calculate the thermoelastic-creep response. Calculated stresses are compared with a brittle-fracture stress criterion. An independent computer code is used to calculate fission-gas-induced fuel swelling. Results obtained with the computer simulation indicate that swelling is not a mode of initial fuel deformation. Faster transients result in fuel melting, while slower transients result in fuel cracking. For investigated faster coolant flow coastdowns with time constants of 1 second and 10 seconds, compressive stresses in the outer radial portion of the fuel limit fuel swelling and inhibit fuel cracking. For a slower coolant flow coastdown with a 300 second time constant, tensile stresses in the outer radial portion of the fuel induce early fuel cracking before any melting or significant fuel swelling has occurred. Suggestions for further research are discussed. A derived noniterative solution for mechanics calculations may offer an order of magnitude decrease in computational effort

Correlates of Inconsistent Refusal of Unprotected Sex among Armenian Female Sex Workers

Directory of Open Access Journals (Sweden)

Karine Markosyan

2014-01-01

Full Text Available This cross-sectional study assessed the prevalence and correlates of inconsistent refusal of unprotected sex among female sex workers (FSWs in Armenia. One hundred and eighteen street-based FSWs between the ages of 20 and 52 completed a questionnaire assessing FSWs’ demographic, psychosocial, and behavioral characteristics. A total of 52.5% (n=62 of FSWs reported inconsistent refusal of unprotected sex with clients in the past 3 months. Logistic regression analysis controlling for participants’ age and education revealed that perceiving more barriers toward condom use (AOR = 1.1; P<0.01, reporting more types of abuse (AOR = 2.1; P<0.01, and setting lower fees for service (AOR = 0.9; P=0.02 significantly predicted inconsistent refusal of unprotected sex. HIV-risk-reduction behavioral interventions tailored to FSWs working in Yerevan Armenia should address the factors identified in this study toward the goal of enhancing refusal of unprotected sex and ultimately preventing acquisition of sexually transmitted infections (STIs including HIV.

Transient hair loss in patients with chronic spontaneous urticaria treated with omalizumab

DEFF Research Database (Denmark)

Noshela Ghazanfar, M; Thomsen, S F

2017-01-01

Summary: Omalizumab (anti-IgE) is used as add-on therapy for antihistamine refractory chronic urticaria patients. The most commonly reported adverse effects were headache, arthralgia, upper respiratory infections, fatigue, nausea and injection-site reactions. However, lately a few cases of hair...... loss have been reported. We describe a case of transient hair loss in a young female patient after initiating treatment with omalizumab. Despite this side effect, the patient continued with omalizumab treatment for 10 months with good effect....

Implementation into a CFD code of neutron kinetics and fuel pin models for nuclear reactor transient analyses

International Nuclear Information System (INIS)

Chen Zhao; Chen, Xue-Nong; Rineiski, Andrei; Zhao Pengcheng; Chen Hongli

2014-01-01

Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)

Loss of integrated control system power and overcooling transient at Rancho Seco on December 26, 1985

International Nuclear Information System (INIS)

1986-02-01

On December 26, 1985, Rancho Seco Nuclear Generating Station, located in Clay, California, about 25 miles southeast of Sacramento, experienced a loss of dc power within the integrated control system (ICS) while the plant was operating at 76% power. The plant is owned by the Sacramento Municipal Utility District (SMUD). Following the loss of ICS dc power, the reactor tripped on high reactor coolant system (RCS) pressure followed by a rapid overcooling transient and automatic initiation of the safety features actuation system on low RCS pressure. The overcooling transient continued until ICS dc power was restored 26 minutes after its loss. The fundamental causes for this transient were design weaknesses and vulnerabilities in the ICS and in the equipment controlled by that system. These weaknesses and vulnerabilities were not adequately compensated by other design features, plant procedures or operator training. These weaknesses and vulnerabilities were largely known to SMUD and the NRC staff by virtue of a number of precursor events and through related analyses and studies. Yet, adequate plant modifications were not made so that this event would be improbable, or so that its course or consequences would be altered significantly. The information was available and known which could have prevented this overcooling transient; but in the absence of adequate plant modifications, the incident should have been expected. The report includes findings and conclusions of the NRC Incident Investigation Team sent to Rancho Seco by the NRC Executive Director for Operations in conformance with NRC's recently established Incident Investigation Program. 33 figs

The calculation of dryout during flow and pressure transients

International Nuclear Information System (INIS)

James, P.W.; Whalley, P.B.

1981-01-01

The method for predicting dryout in a round tube by means of an annular flow model (Whalley et al 1974) is extended to cover the case where both inlet mass flux and pressure are time-dependent. The qualitative effects of an inlet pressure transient are assessed by performing a 'numerical experiment' and it is found that the predictions of the model represent reasonable physical trends. The relative merits of wo numerical solution schemes are also discussed

Characterization of transient groundwater flow through a high arch dam foundation during reservoir impounding

Directory of Open Access Journals (Sweden)

Yifeng Chen

2016-08-01

Full Text Available Even though a large number of large-scale arch dams with height larger than 200 m have been built in the world, the transient groundwater flow behaviors and the seepage control effects in the dam foundations under difficult geological conditions are rarely reported. This paper presents a case study on the transient groundwater flow behaviors in the rock foundation of Jinping I double-curvature arch dam, the world's highest dam of this type to date that has been completed. Taking into account the geological settings at the site, an inverse modeling technique utilizing the time series measurements of both hydraulic head and discharge was adopted to back-calculate the permeability of the foundation rocks, which effectively improves the uniqueness and reliability of the inverse modeling results. The transient seepage flow in the dam foundation during the reservoir impounding was then modeled with a parabolic variational inequality (PVI method. The distribution of pore water pressure, the amount of leakage, and the performance of the seepage control system in the dam foundation during the entire impounding process were finally illustrated with the numerical results.

Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

International Nuclear Information System (INIS)

Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

1985-04-01

The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

Effect of vortex generators on the closing transient flow of bileaflet mechanical heart valves

Science.gov (United States)

Murphy, David; Dasi, Lakshmi; Yoganathan, Ajit; Glezer, Ari

2006-11-01

The time-periodic closing of bileaflet mechanical heart valves is accompanied by a strong flow transient that is associated with the formation of a counter-rotating vortex pair near the b-datum line of leaflet edges. The strong transitory shear that is generated by these vortices may be damaging to blood elements and may result in platelet activation. In the present work, these flow transients are mitigated using miniature vortex generator arrays that are embedded on the surface of the leaflets. Two vortex generator designs were investigated: one design comprised staggered rectangular fins and the other one staggered hemispheres. The closing transients in the absence and presence of the passive vortex generators are characterized using phase locked PIV measurements. The study utilizes a 25 mm St. Jude Medical valve placed in the aortic position of the Georgia Tech left heart simulator. Measurements of the velocity field in the center plane of the leaflets demonstrate that the dynamics of the transient vortices that precede the formation of the leakage jets can be significantly altered and controlled by relatively simple passive modifications of existing valve designs. Human blood experiments validated the effectiveness of miniature vortex generators in reducing thrombus formation by over 42 percent.

Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2005-01-01

Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

International Nuclear Information System (INIS)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Aronne, Ivan D.; Rezende, Guilherme P.

2011-01-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F., E-mail: antonella@nuclear.ufmg.br, E-mail: laubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, UFMG, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores, CNPq (Brazil); Aronne, Ivan D.; Rezende, Guilherme P., E-mail: aroneid@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil).

2011-07-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

International Nuclear Information System (INIS)

Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

1983-01-01

Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

Structural evaluation of electrosleeved tubes under severe accident transients

International Nuclear Information System (INIS)

Majumdar, S.

1999-01-01

A flow stress model was developed for predicting failure of Electrosleeved PWR steam generator tubing under severe accident transients. The Electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400 C during severe accidents because of grain growth. A grain growth model and the Hall-Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data as well as high temperature failure tests on notched Electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of Electrosleeved tubes with axial cracks in the parent tube during postulated severe accident transients

Status of the Astrid core at the end of the pre-conceptual design phase 1

International Nuclear Information System (INIS)

Chenaud, Ms.; Devictor, N.; Mignot, G.; Varaine, F.; Venard, C.; Martin, L.; Phelip, M.; Lorenzo, D.; Serre, F.; Bertrand, F.; Alpy, N.; Le-Flem, M.; Gavoille, P.; Lavastre, R.; Richard, P.; Verrier, D.; Schmitt, D.

2013-01-01

Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase. (authors)

Transient Stefan flow and thermophoresis around an evaporating droplet

International Nuclear Information System (INIS)

Vittori, O.

1984-01-01

The particle scavening efficiency of vapour-grown ice crystals falling from mixed clouds proves to be very high. Stefan flow, an aerodynamic flow originating in the fluid surrounding evaporating or condensing bodies, pushes airborne particles away from the surface of the supercooled droplets evaporating in the vicinity of an ice crystal. The particle Brownian flux towards the surface of the ice crystal (terminal velocity of about 1 m s -1 ) is, therefore, enhanced. However, the efficiency of this process of airborne-particle removal is strongly reduced as a consequence of the cooling of the evaporating droplet which produces a ''thermal force'', thermophoresis, which counteracts the particle Stefan flow. At the surface of an evaporating droplet in a quasi-equilibrium state, the two airborne-particle velocity fields practically balance each other. This counteracting effect on particle motion needs to be evaluated in the transient case. An approach is presented which consists of reformulating the transient heat and mass transfer problem in such a way as to convert it into a purely heat transfer problem having a known analytical solution. The approach is discussed and found to be correct. The results of the computations show that the counteracting role of thermophoresis on Stefan-flow particle motion during the residence time of supercooled droplets in the vicinity of an ice crystal (from 10 -5 to 10 -4 s), which is also the time in which evaporation takes place, is considerably weak. It turns out to be practically negligible for large droplets (radius >= 8x10 -4 cm)

Analysis of cell flow and cell loss following X-irradiation using sequential investigation of the total number of cells in the various parts of the cell cycle

International Nuclear Information System (INIS)

Skog, S.; Tribukait, B.

1985-01-01

The cell flow and cell loss of an in vivo growing Ehrlich ascites tumour were calculated by sequential estimation of changes in total number of cells in the cell cycle compartments. Normal growth was compared with the grossly disturbed cell flow evident after a 5 Gy X-irradiation. The doubling time of normal, exponentially growing cells was 24 hr. The generation time was 21 hr and the potential doubling time was 21 hr. Thus, the growth fraction was 1.0 and the cell loss rate about 0.5%/hr. Following irradiation, a transiently increased relative outflow rate from all cell cycle compartments was found at about 3 and 40 hr, and from S phase at 24 hr after irradiation. Increase in cell loss as well as non-viable cells was observed at 24 hr after irradiation at the time of release of the irradiation-induced G 2 blockage. The experiments show the applicability and limitations of cell flow and cell loss calculations by sequential analysis of the total number of cells in the various parts of the cell cycle. (author)

CFD simulating the transient thermal–hydraulic characteristics in a 17 × 17 bundle for a spent fuel pool under the loss of external cooling system accident

International Nuclear Information System (INIS)

Chen, S.R.; Lin, W.C.; Ferng, Y.M.; Chieng, C.C.; Pei, B.S.

2014-01-01

Highlights: • A 3-D CFD is adopted to simulate transient behaviors in an SFP under the accident. • This model realistically simulates a 17 × 17 bundle, rid of porous media approach. • The loss of external cooling system accident for an SFP is assumed in this paper. • Thermal–hydraulic characteristics in a bundle are strongly influenced by grids. • The results confirm temperature rising rate used in Maanshan NPP is conservative. - Abstract: This paper develops a three-dimensional (3-D) transient computational fluid dynamics (CFD) model to simulate the thermal–hydraulic characteristics in a fuel bundle located in a spent fuel pool (SFP) under the loss of external cooling system accident. The SFP located in the Maanshan nuclear power plant (NPP) is selected herein. Without adopting the porous media approach usually used in the previous CFD works, this model uses a real-geometry simulation of a 17 × 17 fuel bundle, which can obtain the localized distributions of the flow and heat transfer during the accident. These distribution characteristics include several peaks in the axial distributions of flow, pressure, temperature, and Nusselt number (Nu) near the support grids, the non-uniform distribution of secondary flow, and the non-uniform temperature distribution due to flow mixing between rods, etc. According to the conditions adopted in the Procedure 597.1 (MNPP Plant Procedure 597.1, 2010) for the management of the loss-of-cooling event of the spent fuel pool in the Maanshan NPP, the temperature rising rate predicted by the present model can be equivalent to 1.26 K/h, which is the same order as that of 3.5 K/h in the this procedure. This result also confirms that the temperature rising rate used in the Procedure 597.1 for the Maanshan NPP is conservative. In addition, after the loss of external cooling system, there are about 44 h for the operator to repair the malfunctioning system or provide the alternative water source for the pool inventory to

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

Energy Technology Data Exchange (ETDEWEB)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

Investigation of combined free and forced convection in a 2 x 6 rod bundle during controlled flow transients

International Nuclear Information System (INIS)

Bates, J.M.; Khan, E.U.

1980-10-01

An experimental study was performed to obtain local fluid velocity and temperature measurements in the mixed (combined free and forced) convection regime for specific flow coastdown transients. A brief investigation of steady-state flows for the purely free-convection regime was also completed. The study was performed using an electrically heated 2 x 6 rod bundle contained in a flow housing. In addition a transient data base was obtained for evaluating the COBRA-WC thermal-hydraulic computer program

Transient flows in rectangular MHD ducts under the influence of suddenly changing applied magnetic fields

International Nuclear Information System (INIS)

Kobayashi, Junichi

1979-01-01

The study on the transient flow characteristics in MHD ducts under orthogonal magnetic field is divided into handling two problems: the problem of changing pressure gradient in a uniform orthogonal magnetic field and the problem in which the orthogonal magnetic field itself changes with time. The former has been investigated by many persons, but the latter has not been investigated so often as the former because of its difficulty of handling. In addition, if it is intended to grasp properly the transient flow characteristics in actual MHD ducts, it will be also important that the effects of the electric conductivity of side walls and aspect ratio are clarified. In other words, this paper deals with the problem in which a uniform orthogonal magnetic field is suddenly applied in such manner as Heaviside's step function to or removed from the conductive fluids flowing in sufficiently long rectangular MHD ducts. First, the MHD fundamental equations are described, then they are normalized to give boundary conditions and initial conditions. Next, the transient flow and the derived magnetic field characteristics are numerically analyzed by the difference calculus, and thus the effects of conductor, insulated wall, aspect ratio, Hartmann number, magnetic Prandtl number and others on the above characteristics are clarified. (Wakatsuki, Y.)

Transient Analysis of Generation IV quick reactors; Analisis de Transitorios en Reactores Rapidos de Generacion IV

Energy Technology Data Exchange (ETDEWEB)

Vazquez, M.; Martin-Fuertes, F.

2013-07-01

As a complement to the attached code 3D neutron-CIEMAT thermohydraulic added a module to simulate transient. Temporary kinetics is resolved by factoring flow in a spatial part and another storm. MCNP provides the reactivity and updated spatial function and COBRA-IV calculates the temperature distribution. Temporary dependence of amplitude is calculated using time delayed neutron Kinetic equations. As an example of application, examines a transient loss of flow in MYRRHA, a lead-cooled experimental reactor.

Numerical methods and transition investigation of transient flows around a pitching hydrofoil

International Nuclear Information System (INIS)

Wu, Q; Wang, G Y; Huang, B

2013-01-01

The numerical simulations for a NACA66 hydrofoil are performed by using the standard k-ω SST turbulence model and revised γ-Re θ transition model respectively. The simulation results are compared with the experimental results, and the hydrodynamic property and the fluid structure during the pitching process is studied. It is revealed that, compared with the standard k-ω SST turbulence model, the revised γ-Re θ transition model is able to present the hydrodynamic property and the fluid structure of the transient flow around a pitching hydrofoil more accurately, and better predict the separation and transition process in the boundary layer. The transient flow process around a pitching hydrofoil can be divided into 5 parts. At small angle of attack, transition is observed at the leading edge of the foil, resulting in the inflection of dynamic property curves. As the angle of attack increases, a clockwise trailing edge vortex expands toward the leading edge of the foil. At high angles of attack, large-scale load fluctuations are observed due to the stall caused by separation of the leading edge vortex. The flow transitions back to laminar during the downward pitching process

TRAC-PF1 analysis of LOFT steam-generator feedwater transient test L9-1

International Nuclear Information System (INIS)

Meier, J.K.

1983-01-01

The Transient Reactor Analysis Code (TRAC-PF1) calculations were compared to test data from Loss-of-Fluid Test (LOFT) L9-1, which was a loss-of-feedwater transient. This paper includes descriptions of the test and the TRAC input and compares the TRAC-calculated results with the test data. We conclude that the code predicted the experiment well, given the uncertainties in the boundary conditions. The analysis indicates the need to model all the flow paths and heat structures, and to improve the TRAC wall condensation heat-transfer model

Experimental Setup with Transient Behavior of Fuel Cladding of SFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Sang Hun; Kim, Jun Hwan; Kim, June-Hyung; Ryu, Woo Seog; Park, Sang Gyu; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Nowadays, in Korea, advanced cladding such as FC92 is developed and its transient behaviors are required for the safety analysis of SFR. Design and safety analyses of sodium-cooled fast reactor (SFR) require understanding fuel pin responses to a wide range of off-normal events. In a loss-of-flow (LOF) or transient over-power (TOP), the temperature of the cladding is rapidly increased above its steady-state service temperature. Transient tests have been performed in sections of fuel pin cladding and a large data base has been established for austenitic stainless steel such as 20% cold-worked 316 SS and ferritic/martensitic steels such as HT9. This paper summarizes the technical status of transient testing facilities and their results. Previous researches showed the transient behaviors of HT9 cladding. For the safety analyses in SFR in Korea, simulated transient tests with newly developed FC92 as well as HT9 cladding are being carried out.

Transient heat transfer for helium gas flowing over a horizontal cylinder with exponentially increasing heat input

International Nuclear Information System (INIS)

Liu, Qiusheng; Fukuda, Katsuya

2003-01-01

The transient heat transfer coefficients for forced convection flow of helium gas over a horizontal cylinder were measured under wide experimental conditions. The platinum cylinder with a diameter of 1.0 mm was used as test heater and heated by electric current with an exponentially increasing heat input of Q 0 exp(t/τ). The gas flow velocities ranged from 5 to 35 m/s, the gas temperatures ranged from 25 to 80degC, and the periods of heat generation rate, τ, ranged from 40 ms to 20 s. The surface superheat and heat flux increase exponentially as the heat generation rate increases with the exponential function. It was clarified that the heat transfer coefficient approaches the quasi-steady-state one for the period τ longer than about 1 s, and it becomes higher for the period shorter than around 1 s. The transient heat transfer shows less dependence on the gas flowing velocity when the period becomes very shorter. The gas temperature in this study shows little influence on the heat transfer coefficient. Semi-empirical correlation for quasi-steady-state heat transfer was obtained based on the experimental data. The ratios of transient Nusselt number Nu tr to quasi-steady-state Nusselt number Nu st at various periods, flow velocities, and gas temperatures were obtained. The heat transfer shifts to the quasi-steady-state heat transfer for longer periods and shifts to the transient heat transfer for shorter periods at the same flow velocity. It also approaches the quasi-steady-state one for higher flow velocity at the same period. Empirical correlation for transient heat transfer was also obtained based on the experimental data. (author)

Loss of flow accident and its mitigation measures for nuclear systems with SCWR-M

International Nuclear Information System (INIS)

Xu Zhihong; Hou Dong; Fu Shengwei; Yang Yanhua; Cheng Xu

2011-01-01

Highlights: → A model of mixed spectrum SCWR system is established by a revised version of RELAP5. → Some important parameters are chosen to analysis the SCWR-M during LOFA. → Three important mitigation measures for LOFA of SCWR-M are derived from the results. - Abstract: Based on a revised version of RELAP5, which can be used for super-critical pressure calculation, a model of mixed spectrum SCWR (SCWR-M) system is established. To analyze the transient behavior of SCWR-M and develop mitigation measures during loss of flow accident (LOFA), some important parameters, e.g. reactor coolant pump (RCP) coast-down time, Reactor Pressure Vessel (RPV) upper water volume and safety injection flow, etc., are chosen for the parametric analysis. The results achieved so far indicate that the SCWR-M system design is feasible and promising. Three important mitigation measures for LOFA of SCWR-M are derived from the results: RCP coast-down time of more than 15 s, RPV upper water volume of more than 27 m 3 , and safety injection of more than 5% of the system design flow.

Flow-excursion-induced dryout at low-heat-flux

International Nuclear Information System (INIS)

Khatib-Rahbar, M.; Cazzoli, E.G.

1983-01-01

Flow-excursion-induced dryout at low-heat-flux natural-convection boiling, typical of liquid-metal fast-breeder reactors, is addressed. Steady-state calculations indicate that low-quality boiling is possible up to the point of Ledinegg instability leading to flow excursion and subsequent dryout in agreement with experimental data. A flow-regime-dependent dryout heat flux relationship based upon saturated boiling criterion is also presented. Transient analysis indicates that premature flow excursion can not be ruled out and sodium boiling is highly transient dependent. Analysis of a high-heat-flux forced convection, loss-of-flow transient shows a significantly faster flow excursion leading to dryout in excellent agreement with parallel calculations using the two-dimensional THORAX code. 17 figures

ERP-IV-A program for transient thermal-hydraulic analysis of PWR plant

International Nuclear Information System (INIS)

Dai Anguo; Tang Jiahuan; Qian Huifu; Gao Zhikang

1987-12-01

The author deal with the descriptions of physical model of transient process in PWR plant and the function of ERP-IV (ERR-IV Transient Thermo-Hydraulic Analysis Code). The code has been developed for safety analysis and design transient. The code is characterized by the multi-loop long-term, short term, wide-range plant simulation with the capability to analyze natural circulation condition. The description of ERP-IV includes following parts: reactor, primary coolant loops, pressurizer, steam generators, main steam system, turbine, feedwater system, steam dump, relive valves, and safety valves in secondary side, etc.. The code can use for accident analysis, such as loss of all A.C. power to power plant auxiliaries (a station blackout), loss of normal feedwater, loss of load, loss of condenser vacuum and other events causing a turbine trip, complete loss of forced reactor coolant flow, uncontrolled rod cluster control assembly bank withdrawal. It can also be used for accident analysis of the emergency and limiting conditions, such as feedwater line break and main steam line rupture. It can also be utilized as a tool for system design studies, component design, setpoint studies and design transition studies, etc

Death Valley regional ground-water flow system, Nevada and California -- hydrogeologic framework and transient ground-water flow model

Science.gov (United States)

Belcher, Wayne R.

2004-01-01

A numerical three-dimensional (3D) transient ground-water flow model of the Death Valley region was developed by the U.S. Geological Survey for the U.S. Department of Energy programs at the Nevada Test Site and at Yucca Mountain, Nevada. Decades of study of aspects of the ground-water flow system and previous less extensive ground-water flow models were incorporated and reevaluated together with new data to provide greater detail for the complex, digital model. A 3D digital hydrogeologic framework model (HFM) was developed from digital elevation models, geologic maps, borehole information, geologic and hydrogeologic cross sections, and other 3D models to represent the geometry of the hydrogeologic units (HGUs). Structural features, such as faults and fractures, that affect ground-water flow also were added. The HFM represents Precambrian and Paleozoic crystalline and sedimentary rocks, Mesozoic sedimentary rocks, Mesozoic to Cenozoic intrusive rocks, Cenozoic volcanic tuffs and lavas, and late Cenozoic sedimentary deposits of the Death Valley Regional Ground-Water Flow System (DVRFS) region in 27 HGUs. Information from a series of investigations was compiled to conceptualize and quantify hydrologic components of the ground-water flow system within the DVRFS model domain and to provide hydraulic-property and head-observation data used in the calibration of the transient-flow model. These studies reevaluated natural ground-water discharge occurring through evapotranspiration and spring flow; the history of ground-water pumping from 1913 through 1998; ground-water recharge simulated as net infiltration; model boundary inflows and outflows based on regional hydraulic gradients and water budgets of surrounding areas; hydraulic conductivity and its relation to depth; and water levels appropriate for regional simulation of prepumped and pumped conditions within the DVRFS model domain. Simulation results appropriate for the regional extent and scale of the model were

Visualizing the transient electroosmotic flow and measuring the zeta potential of microchannels with a micro-PIV technique.

Science.gov (United States)

Yan, Deguang; Nguyen, Nam-Trung; Yang, Chun; Huang, Xiaoyang

2006-01-14

We have demonstrated a transient micro particle image velocimetry (micro-PIV) technique to measure the temporal development of electroosmotic flow in microchannels. Synchronization of different trigger signals for the laser, the CCD camera, and the high-voltage switch makes this measurement possible with a conventional micro-PIV setup. Using the transient micro-PIV technique, we have further proposed a method on the basis of inertial decoupling between the particle electrophoretic motion and the fluid electroosmotic flow to determine the electrophoretic component in the particle velocity and the zeta potential of the channel wall. It is shown that using the measured zeta potentials, the theoretical predictions agree well with the transient response of the electroosmotic velocities measured in this work.

Development of a computer code for Dalat research reactor transient analysis

International Nuclear Information System (INIS)

Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong

2003-01-01

DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)

On the potential importance of transient air flow in advective radon entry into buildings

International Nuclear Information System (INIS)

Narasimhan, T.N.; Tsang, Y.W.; Holman, H.Y.

1990-01-01

The authors have investigated, using a mathematical model, the temporal variations of air flux within the soil mass surrounding a basement in the presence of time dependent periodic variations of barometric pressure and a persistent under-pressure at the basement. The results of transient air flow show that for a homogeneous soil medium, the effects of barometric fluctuations are most significant in the cases where soil permeability to air is low and the fluctuation frequency is high. In these cases, the barometric fluctuation can greatly enhance the magnitude of fluxes as well as introduce flow direction reversals from surrounding soil into the basement. These large fluxes with direction reversals have strong implications in regard to advective transport of radon. The results suggest that the transient oscillations have to be accounted for in quantifying radon entry into buildings. In the actual field set up, the transient behavior will be further influenced by soil permeability heterogeneity, by soil moisture variations, and by the effects of multiple periodic components in the barometric pressure fluctuations

Experiment for transient effects of sudden catastrophic loss of vacuum on a scaled superconducting radio frequency cryomodule

International Nuclear Information System (INIS)

Dalesandro, A.; Theilacker, J.; Van Sciver, S.W.

2011-01-01

Safe operation of superconducting radio frequency (SRF) cavities require design consideration of a sudden catastrophic loss of vacuum (SCLV) adjacent with liquid helium (LHe) vessels and subsequent dangers. An experiment is discussed to test the longitudinal effects of SCLV along the beam line of a string of scaled SRF cavities. Each scaled cavity includes one segment of beam tube within a LHe vessel containing 2 K saturated LHe, and a riser pipe connecting the LHe vessel to a common gas header. At the beam tube inlet is a fast acting solenoid valve to simulate SCLV and a high/low range orifice plate flow-meter to measure air influx to the cavity. The gas header exit also has an orifice plate flow-meter to measure helium venting the system at the relief pressure of 0.4 MPa. Each cavity is instrumented with Validyne pressure transducers and Cernox thermometers. The purpose of this experiment is to quantify the time required to spoil the beam vacuum and the effects of transient heat and mass transfer on the helium system. Heat transfer data is expected to reveal a longitudinal effect due to the geometry of the experiment. Details of the experimental design criteria and objectives are presented.

Design and safety studies on an EFIT core with CERMET fuel

International Nuclear Information System (INIS)

Chen, Xue-Nong; Rineiski, Andrei; Liu, Ping; Maschek, Werner; Matzerath Boccaccini, Claudia; Gabrielli, Fabrizio; Sobolev, Vitaly

2008-01-01

Within the EUROTRANS Programme a European Facility for Industrial Transmutation (EFIT) is under development. This paper deals with the design and safety analyses of an EFIT core with Mo-matrix based CERMET fuel. A three zone core design was developed, which satisfies the EFIT general and specific requirements. The fuel/matrix ratio in each zone is determined for a suitable subcritical level at a k eff of about 0.97 and a total form factor around 1.5. The Pu/MA ratio also determines the transmutation rate and the burn-up characteristics, ranging between 46/54 at% to 40/60 at% for optimizing the reactivity swing and the MA transmutation efficiency. Based on the preliminary core design, safety calculations are performed with SIMMER-III for three types of transient: the unprotected loss of flow (ULOF), the unprotected transient of over power (UTOP) and the unprotected blockage accident (UBA). It can be shown that in the CERMET core the fuel and clad design limits are not violated under the conditions of ULOF and UTOP. In the UBA case, pin failures will happen and lead to a local voiding and reactivity insertion, but a fuel sweep-out process leads to a power reduction and restricts the core degradation. (authors)

Direct chemoselective synthesis of glyconanoparticles from unprotected reducing glycans and glycopeptide aldehydes

DEFF Research Database (Denmark)

Thygesen, Mikkel Boas; Sørensen, Kasper Kildegaard; Cló, Emiliano

2009-01-01

Chemoselective oxime coupling was used for facile conjugation of unprotected, reducing glycans and glycopeptide aldehydes with core-shell gold nanoparticles carrying reactive aminooxy groups on the organic shell.......Chemoselective oxime coupling was used for facile conjugation of unprotected, reducing glycans and glycopeptide aldehydes with core-shell gold nanoparticles carrying reactive aminooxy groups on the organic shell....

Transmutation of Americium in Fast Neutron Facilities

International Nuclear Information System (INIS)

Zhang, Youpeng

2011-01-01

In this thesis, the feasibility to use a medium sized sodium cooled fast reactor fully loaded with MOX fuel for efficient transmutation of americium is investigated by simulating the safety performance of a BN600-type fast reactor loaded with different fractions of americium in the fuel, using the safety parameters obtained with the SERPENT Monte Carlo code. The focus is on americium mainly due to its long-term contribution to the radiotoxicity of spent nuclear fuel and its deterioration on core's safety parameters. Applying the SAS4A/SASSYS transient analysis code, it is demonstrated that the power rating needs to be reduced by 6% for each percent additional americium introduction into the reference MOX fuel, maintaining 100 K margin to fuel melting, which is the most limiting failure mechanism. Safety analysis of a new Accelerator Driven System design with a smaller pin pitch-to-diameter ratio comparing to the reference EFIT-400 design, aiming at improving neutron source efficiency, was also performed by simulating performance for unprotected loss of flow, unprotected transient overpower, and protected loss-of-heat-sink transients, using neutronic parameters obtained from MCNP calculations. Thanks to the introduction of the austenitic 15/15Ti stainless steel with enhanced creep rupture resistance and acceptable irradiation swelling rate, the suggested ADS design loaded with nitride fuel and cooled by lead-bismuth eutectic could survive the full set of transients, preserving a margin of 130 K to cladding rupture during the most limiting transient. The thesis concludes that efficient transmutation of americium in a medium sized sodium cooled fast reactor loaded with MOX fuel is possible but leads to a severe power penalty. Instead, preserving transmutation rates of minor actinides up to 42 kg/TWh th , the suggested ADS design with enhanced proton source efficiency appears like a better option for americium transmutation

Intuitive Visualization of Transient Flow: Towards a Full 3D Tool

Science.gov (United States)

Michel, Isabel; Schröder, Simon; Seidel, Torsten; König, Christoph

2015-04-01

Visualization of geoscientific data is a challenging task especially when targeting a non-professional audience. In particular, the graphical presentation of transient vector data can be a significant problem. With STRING Fraunhofer ITWM (Kaiserslautern, Germany) in collaboration with delta h Ingenieurgesellschaft mbH (Witten, Germany) developed a commercial software for intuitive 2D visualization of 3D flow problems. Through the intuitive character of the visualization experts can more easily transport their findings to non-professional audiences. In STRING pathlets moving with the flow provide an intuition of velocity and direction of both steady-state and transient flow fields. The visualization concept is based on the Lagrangian view of the flow which means that the pathlets' movement is along the direction given by pathlines. In order to capture every detail of the flow an advanced method for intelligent, time-dependent seeding of the pathlets is implemented based on ideas of the Finite Pointset Method (FPM) originally conceived at and continuously developed by Fraunhofer ITWM. Furthermore, by the same method pathlets are removed during the visualization to avoid visual cluttering. Additional scalar flow attributes, for example concentration or potential, can either be mapped directly to the pathlets or displayed in the background of the pathlets on the 2D visualization plane. The extensive capabilities of STRING are demonstrated with the help of different applications in groundwater modeling. We will discuss the strengths and current restrictions of STRING which have surfaced during daily use of the software, for example by delta h. Although the software focusses on the graphical presentation of flow data for non-professional audiences its intuitive visualization has also proven useful to experts when investigating details of flow fields. Due to the popular reception of STRING and its limitation to 2D, the need arises for the extension to a full 3D tool

Transient behaviour of main coolant pump in nuclear power plants

International Nuclear Information System (INIS)

Delja, A.

1986-01-01

A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

Characteristics of Nitrogen Loss through Surface-Subsurface Flow on Red Soil Slopes of Southeast China

Science.gov (United States)

Zheng, Haijin; Liu, Zhao; Zuo, Jichao; Wang, Lingyun; Nie, Xiaofei

2017-12-01

Soil nitrogen (N) loss related to surface flow and subsurface flow (including interflow and groundwater flow) from slope lands is a global issue. A lysimetric experiment with three types of land cover (grass cover, GC; litter cover, LC; and bare land, BL) were carried out on a red soil slope land in southeast China. Total Nitrogen (TN) loss through surface flow, interflow and groundwater flow was observed under 28 natural precipitation events from 2015 to 2016. TN concentrations from subsurface flow on BL and LC plots were, on average, 2.7-8.2 and 1.5-4.4 times greater than TN concentrations from surface flow, respectively; the average concentration of TN from subsurface flow on GC was about 36-56% of that recorded from surface flow. Surface flow, interflow and groundwater flow contributed 0-15, 2-9 and 76-96%, respectively, of loss load of TN. Compared with BL, GC and LC intercepted 83-86% of TN loss through surface runoff; GC intercepted 95% of TN loss through subsurface flow while TN loss through subsurface flow on LC is 2.3 times larger than that on BL. In conclusion, subsurface flow especially groundwater flow is the dominant hydrological rout for N loss that is usually underestimated. Grass cover has the high retention of N runoff loss while litter mulch will increase N leaching loss. These findings provide scientific support to control N runoff loss from the red soil slope lands by using suitable vegetation cover and mulching techniques.

Manual for operation of the multipurpose thermalhydraulic test facility TOPFLOW (Transient Two Phase Flow Test Facility)

International Nuclear Information System (INIS)

Beyer, M.; Carl, H.; Schuetz, H.; Pietruske, H.; Lenk, S.

2004-07-01

The Forschungszentrum Rossendorf (FZR) e. V. is constructing a new large-scale test facility, TOPFLOW, for thermalhydraulic single effect tests. The acronym stands for transient two phase flow test facility. It will mainly be used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of computational fluid dynamic (CFD) codes. The manual of the test facility must always be available for the staff in the control room and is restricted condition during operation of personnel and also reconstruction of the facility. (orig./GL)

Development of the computer code for transient analysis in experimental fast reactor

International Nuclear Information System (INIS)

Moreira, M.L.; Sato, E.F.

1989-01-01

A calculational model of heat transfer and fluid coolant dynamics, for thermal-hydraulic simulation of the primary system components of a pool type experimental fast breeder reactor, has developed. Programmed in FORTRAN, the SORES code was used to simulate transients as loss of pumping and loss of secondary sodium flow in the EBRII. The SORES results compared with measured data and NATDEMO code results was found to be good. (author) [pt

The effect of the virtual mass force term on the stability of transient two-phase flow analysis

International Nuclear Information System (INIS)

Watanabe, Tadashi; Hirano, Masashi; Tanabe, Fumiya

1989-08-01

The effect of the virtual mass force term on the stability of transient two-phase flow analysis is studied. The objective form of the virtual mass acceleration is used. The virtual mass coefficient is determined from the stability condition of basic equations against infinitesimal high wave-number perturbations. The parameter is chosen so that a reasonable agreement between the analytical and experimental sound speed in two-phase flows can be obtained. A one-dimensional sedimentation problem is simulated by the MINCS code which is a tool for transient two-phase flow analysis. The stability analysis is performed for the numerical procedure. It is shown that calculated results are stabilized so long as the virtual mass coefficient satisfies the stability condition of differential equations. (author)

Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

2009-06-15

Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

MINET, Transient Fluid Flow and Heat Transfer Power Plant Network Analysis

International Nuclear Information System (INIS)

Van Tuyle, G.J.

2002-01-01

1 - Description of program or function: MINET (Momentum Integral Network) was developed for the transient analysis of intricate fluid flow and heat transfer networks, such as those found in the balance of plant in power generating facilities. It can be utilized as a stand-alone program or interfaced to another computer program for concurrent analysis. Through such coupling, a computer code limited by either the lack of required component models or large computational needs can be extended to more fully represent the thermal hydraulic system thereby reducing the need for estimating essential transient boundary conditions. The MINET representation of a system is one or more networks of volumes, segments, and boundaries linked together via heat exchangers only, i.e., heat can transfer between networks, but fluids cannot. Volumes are used to represent tanks or other volume components, as well as locations in the system where significant flow divisions or combinations occur. Segments are composed of one or more pipes, pumps, heat exchangers, turbines, and/or valves each represented by one or more nodes. Boundaries are simply points where the network interfaces with the user or another computer code. Several fluids can be simulated, including water, sodium, NaK, and air. 2 - Method of solution: MINET is based on a momentum integral network method. Calculations are performed at two levels, the network level (volumes) and the segment level. Equations conserving mass and energy are used to calculate pressure and enthalpy within volumes. An integral momentum equation is used to calculate the segment average flow rate. In-segment distributions of mass flow rate and enthalpy are calculated using local equations of mass and energy. The segment pressure is taken to be the linear average of the pressure at both ends. This method uses a two-plus equation representation of the thermal hydraulic behavior of a system of heat exchangers, pumps, pipes, valves, tanks, etc. With the

SASSYS-1 modelling of RVACS/RACS heat removal in an LMR

International Nuclear Information System (INIS)

Dunn, F.E.

1987-01-01

The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This air-side RVACS/RACS model is coupled with the sodium-side primary loop thermal hydraulics model in SASSYS-1 to give a complete treatment of the problem. Application of this model to an unprotected loss-of-flow event in the PRISM rector shows that in the long run the RVACS cooling is sufficient to prevent unacceptably high system temperatures in this case

Theoretical and experimental studies on transient heat transfer for forced convection flow of helium gas over a horizontal cylinder

International Nuclear Information System (INIS)

Liu Qiusheng; Katsuya Fukuda; Zhang Zheng

2005-01-01

Forced convection transient heat transfer for helium gas at various periods of exponential increase of heat input to a horizontal cylinder (heater) was theoretically and experimentally studied. In the theoretical study, transient heat transfer was numerically solved based on a turbulent flow model. It was clarified that the surface superheat and heat flux increase exponentially as the heat generation rate increases with the exponential function. The temperature distribution near the cylinder becomes larger as the surface temperature increases. The values of numerical solution for surface temperature and heat flux agree well with the experimental data for the cylinder diameter of 1 mm. However, the heat flux shows difference from the experimental values for the cylinder diameters of 0.7 mm and 2.0 mm. In the experimental studies, the authors measured heat flux, surface temperature, and transient heat transfer coefficients for forced convection flow of helium gas over horizontal cylinders under wide experimental conditions. The platinum cylinders with diameters of 1.0 mm, 0.7 mm, and 2.0 mm were used as test heaters and heated by electric current with an exponential increase of Q 0exp (t/τ) . The gas flow velocities ranged from 2 to 10 m/s, the gas temperatures ranged from 303 to 353 K, and the periods ranged from 50 ms to 20 s. It was clarified that the heat transfer coefficient approaches the quasi-steady-state one for the period τ longer than about 1 s, and it becomes higher for the period shorter than around 1 s. The transient heat transfer shows less dependence on the gas flowing velocity when the period becomes very shorter. The heat transfer shifts to the quasi-steady-state heat transfer for longer periods and shifts to the transient heat transfer for shorter periods at the same flow velocity. It also approaches the quasi-steady-state one for higher flow velocity at the same period. The transient heat transfer coefficients show significant dependence on

ATHENA simulations of divertor loss of heat sink transient for the GSSR - Final report with updates

Energy Technology Data Exchange (ETDEWEB)

Sponton, L.L

2001-05-01

The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that can occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a loss of heat sink at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for this transient have been evaluated and summarised in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. In the current report results from calculations with an updated pressurizer model and pressurizer control system are outlined. The results show that the pressurizer safety valve does not open, that the pressurizer level increase is moderate and that no temperature increases jeopardize the structure integrity.

ATHENA simulations of divertor loss of heat sink transient for the GSSR - Final report with updates

International Nuclear Information System (INIS)

Sponton, L.L.

2001-05-01

The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that can occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a loss of heat sink at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for this transient have been evaluated and summarised in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. In the current report results from calculations with an updated pressurizer model and pressurizer control system are outlined. The results show that the pressurizer safety valve does not open, that the pressurizer level increase is moderate and that no temperature increases jeopardize the structure integrity

Analysis of LOF/TOP excursions in the SNR-300 Zielkern core with a stress controlled rip extension after failure

International Nuclear Information System (INIS)

Royl, P.

1987-07-01

An unprotected loss of flow (ULOF) accident can result in energetic power excursions due to reactivity effects from in-pile fuel motion and compaction after a near midplane failure of fuel pins in unvoided regions of the core (loss of flow driven transient overpower or LOF/TOP excursions). These excursions are strongly mitigated by a propagation of the failure rip, which reduces the fuel compaction and enhances fuel sweep-out. A simple model has been introduced into the SAS3D code in which this propagation is restricted to the expansion of the void zone that is firmed in the coolant channel due to thermal interactions of fuel and sodium (FCI) after failure. The rip propagation is demonstrated in this report in all investigated cases to be a strong mitigator that will limit the fuel compaction effects in the region where they originate

Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

International Nuclear Information System (INIS)

Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao

2008-01-01

This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un-SCRAM-ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role

Segregated copper ratio experiment on transient stability (SeCRETS). Final Report

International Nuclear Information System (INIS)

Bruzzone, P.

2001-01-01

Two Nb 3 Sn, steel jacketed, cable-in-conduit conductors have been manufactured with identical non-Cu cross sections and the stabilizer either included in the Nb 3 Sn composite or partly segregated as copper wires. The two conductors are series connected and wound as a bifilar , single layer solenoid, assembled in the high field bore (11 T) of the SULTAN test facility. The operating current is up to 12 kA (400 A/mm 2 ). A transverse pulsed field is applied with ΔB up to 2.7 T, field rate up to 180 T/s and field integral up to 530 T 2 /s. In the dc test, a good agreement is found between the I c and the T cs results, both correctly scaling according to the parameters derived from the strand tests. The n-value from the V-I curve is in the range of 15. The current sharing at the high field section is correlated with a local current re-distribution, observed by arrays of miniature Hall sensors, detecting the self-field around the conductor. The ac losses results in the range of 2 to 9 Hz by gas flow calorimetry indicate coupling currents constant, nτ, in the range of 1.5 ms at high field, increasing by a factor of 2 with 12 kA transport current. Loss extrapolation to 0 frequency suggests that the loss curve may be not linear outside the test range, with higher nτ at lower field rate. The calorimetric loss estimation at the fast field transient (f=15 Hz) indicates nτ ≅ 2 ms. The ITER plasma disruption transients have been reproduced by the pulsed coils. Due to the very low ac losses, no quench could be generated in either conductor even reducing the temperature margin below 0.2-0.3 K. Very large field transients, with integral above 100 T 2 /s, are required to quench the conductors. In that range, the conductor without segregated copper has superior performance. Due to the large interstrand resistance (very low ac losses), the segregated copper has marginal contribution to the stability. No evidence of current redistribution is observed during the field transients

Sensitivity Analysis of Uncertainty Parameter based on MARS-LMR Code on SHRT-45R of EBR II

Energy Technology Data Exchange (ETDEWEB)

Kang, Seok-Ju; Kang, Doo-Hyuk; Seo, Jae-Seung [System Engineering and Technology Co., Daejeon (Korea, Republic of); Bae, Sung-Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeong, Hae-Yong [Sejong University, Seoul (Korea, Republic of)

2016-10-15

In order to assess the uncertainty quantification of the MARS-LMR code, the code has been improved by modifying the source code to accommodate calculation process required for uncertainty quantification. In the present study, a transient of Unprotected Loss of Flow(ULOF) is selected as typical cases of as Anticipated Transient without Scram(ATWS) which belongs to DEC category. The MARS-LMR input generation for EBR II SHRT-45R and execution works are performed by using the PAPIRUS program. The sensitivity analysis is carried out with Uncertainty Parameter of the MARS-LMR code for EBR-II SHRT-45R. Based on the results of sensitivity analysis, dominant parameters with large sensitivity to FoM are picked out. Dominant parameters selected are closely related to the development process of ULOF event.

CFD Tools for Design and Simulation of Transient Flows in Hypersonic Facilities

Science.gov (United States)

2010-03-24

enthalpy shock tunnel. The Aeronautical Journal, 95(949):324–334, 1991. [6] K. Hannemann , R. Krek, and G. Eitelberg. Latest calibration results of the HEG...K. Hannemann , P. A. Jacobs, J. M. Austin, A. Thomas, and T. J. McIntyre. Transient and steady-state flow in a small shock tube. In A. Paull et al

Transient flows of the solar wind associated with small-scale solar activity in solar minimum

Science.gov (United States)

Slemzin, Vladimir; Veselovsky, Igor; Kuzin, Sergey; Gburek, Szymon; Ulyanov, Artyom; Kirichenko, Alexey; Shugay, Yulia; Goryaev, Farid

The data obtained by the modern high sensitive EUV-XUV telescopes and photometers such as CORONAS-Photon/TESIS and SPHINX, STEREO/EUVI, PROBA2/SWAP, SDO/AIA provide good possibilities for studying small-scale solar activity (SSA), which is supposed to play an important role in heating of the corona and producing transient flows of the solar wind. During the recent unusually weak solar minimum, a large number of SSA events, such as week solar flares, small CMEs and CME-like flows were observed and recorded in the databases of flares (STEREO, SWAP, SPHINX) and CMEs (LASCO, CACTUS). On the other hand, the solar wind data obtained in this period by ACE, Wind, STEREO contain signatures of transient ICME-like structures which have shorter duration (<10h), weaker magnetic field strength (<10 nT) and lower proton temperature than usual ICMEs. To verify the assumption that ICME-like transients may be associated with the SSA events we investigated the number of weak flares of C-class and lower detected by SPHINX in 2009 and STEREO/EUVI in 2010. The flares were classified on temperature and emission measure using the diagnostic means of SPHINX and Hinode/EIS and were confronted with the parameters of the solar wind (velocity, density, ion composition and temperature, magnetic field, pitch angle distribution of the suprathermal electrons). The outflows of plasma associated with the flares were identified by their coronal signatures - CMEs (only in few cases) and dimmings. It was found that the mean parameters of the solar wind projected to the source surface for the times of the studied flares were typical for the ICME-like transients. The results support the suggestion that weak flares can be indicators of sources of transient plasma flows contributing to the slow solar wind at solar minimum, although these flows may be too weak to be considered as separate CMEs and ICMEs. The research leading to these results has received funding from the European Union’s Seventh Programme

Ongoing hydrothermal heat loss from the 1912 ash-flow sheet, Valley of Ten Thousand Smokes, Alaska

Science.gov (United States)

Hogeweg, N.; Keith, T.E.C.; Colvard, E.M.; Ingebritsen, S.E.

2005-01-01

The June 1912 eruption of Novarupta filled nearby glacial valleys on the Alaska Peninsula with ash-flow tuff (ignimbrite), and post-eruption observations of thousands of steaming fumaroles led to the name 'Valley of Ten Thousand Smokes' (VTTS). By the late 1980s most fumarolic activity had ceased, but the discovery of thermal springs in mid-valley in 1987 suggested continued cooling of the ash-flow sheet. Data collected at the mid-valley springs between 1987 and 2001 show a statistically significant correlation between maximum observed chloride (Cl) concentration and temperature. These data also show a statistically significant decline in the maximum Cl concentration. The observed variation in stream chemistry across the sheet strongly implies that most solutes, including Cl, originate within the area of the VTTS occupied by the 1912 deposits. Numerous measurements of Cl flux in the Ukak River just below the ash-flow sheet suggest an ongoing heat loss of ???250 MW. This represents one of the largest hydrothermal heat discharges in North America. Other hydrothermal discharges of comparable magnitude are related to heat obtained from silicic magma bodies at depth, and are quasi-steady on a multidecadal time scale. However, the VTTS hydrothermal flux is not obviously related to a magma body and is clearly declining. Available data provide reasonable boundary and initial conditions for simple transient modeling. Both an analytical, conduction-only model and a numerical model predict large rates of heat loss from the sheet 90 years after deposition.

MVA power flow and loss analysis for electricity market

International Nuclear Information System (INIS)

Wu, Z.Q.; Chen, G.Z.

2001-01-01

MVA power-flow and loss analysis is the basis for allocating the fixed costs and power losses under electricity-market deregulation. It is pointed out that the decomposition allocation of active and reactive power losses is not reasonable. The theory of active and reactive loss allocation and branch-power-flow decomposition has been proposed. Various contributory factors have been deduced. These contributory factors include the contribution factors of the active and reactive generation power, load-power-to-branch flows, the contribution factors of active and reactive generation power to active and reactive load power, the contribution factors of active and reactive load power to generation power, and the contribution factors of active and reactive load power and active and reactive generation power to line power losses. The detailed calculation results are presented and analysed, demonstrating that the theory presented provides a good charging algorithm for all the market participants. (Author)

Simulation of LMFBR pump transients and comparison to LOF that occurred at EBR-II

International Nuclear Information System (INIS)

Koenig, F.F.; Dean, E.M.

1985-01-01

In a large LMFBR plant design, a number of pumps in parallel will feed the core. It must be demonstrated that the plant can continue to operate with the loss of one of the primary pumps. It is desirable not to have check valves in the loop from a reliability and economic standpoint. Simulations have been made to determine the consequences of a loss of one pump in a four-loop pool plant in which no plant protection action is taken. This analysis would be used to determine the required power rundown that would accompany pump loss. The two primary centrifugal pumps in EBR-II feed the core and blanket plenums in two parallel flow paths. The loss of one pump will result in decrease core flow and reverse flow through the down pump since no check valves are present in the system. For a large pool plant with four primary pumps, the loss of one pump will also result in reverse flow through the down pump if check valves of flow diodes are not included. The resulting flow transient has been modeled for EBR-II and the large plant using the DNSP program

Modeling of Transient Nectar Flow in Hummingbird Tongues

Science.gov (United States)

Rico-Guevara, Alejandro; Fan, Tai-Hsi; Rubega, Margaret

2015-11-01

We demonstrate that hummingbirds do not pick up floral nectar via capillary action. The long believed capillary rise models were mistaken and unable to predict the dynamic nectar intake process. Instead, hummingbird's tongue acts as an elastic micropump. Nectar is drawn into the tongue grooves during elastic expansion after the grooves are squeezed flat by the beak. The new model is compared with experimental data from high-speed videos of 18 species and tens of individuals of wild hummingbirds. Self-similarity and transitions of short-to-long time behaviours have been resolved for the nectar flow driven by expansive filling. The transient dynamics is characterized by the relative contributions of negative excess pressure and the apparent area modulus of the tongue grooves.

Analyses of Decrease in Reactor Coolant Flow Rate in SMART

International Nuclear Information System (INIS)

Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

2011-01-01

SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

Transient response in granular quasi-two-dimensional bounded heap flow.

Science.gov (United States)

Xiao, Hongyi; Ottino, Julio M; Lueptow, Richard M; Umbanhowar, Paul B

2017-10-01

We study the transition between steady flows of noncohesive granular materials in quasi-two-dimensional bounded heaps by suddenly changing the feed rate. In both experiments and simulations, the primary feature of the transition is a wedge of flowing particles that propagates downstream over the rising free surface with a wedge front velocity inversely proportional to the square root of time. An additional longer duration transient process continues after the wedge front reaches the downstream wall. The entire transition is well modeled as a moving boundary problem with a diffusionlike equation derived from local mass balance and a local relation between the flux and the surface slope.

Estimation of steady-state and transcient power distributions for the RELAP analyses of the 1963 loss-of-flow and loss-of-pressure tests at BR2

International Nuclear Information System (INIS)

Dionne, B.; Tzanos, C.P.

2011-01-01

To support the safety analyses required for the conversion of the Belgian Reactor 2 (BR2) from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, the simulation of a number of loss-of-flow tests, with or without loss of pressure, has been undertaken. These tests were performed at BR2 in 1963 and used instrumented fuel assemblies (FAs) with thermocouples (TC) imbedded in the cladding as well as probes to measure the FAs power on the basis of their coolant temperature rise. The availability of experimental data for these tests offers an opportunity to better establish the credibility of the RELAP5-3D model and methodology used in the conversion analysis. In order to support the HEU to LEU conversion safety analyses of the BR2 reactor, RELAP simulations of a number of loss-of-flow/loss-of-pressure tests have been undertaken. Preliminary analyses showed that the conservative power distributions used historically in the BR2 RELAP model resulted in a significant overestimation of the peak cladding temperature during the transient. Therefore, it was concluded that better estimates of the steady-state and decay power distributions were needed to accurately predict the cladding temperatures measured during the tests and establish the credibility of the RELAP model and methodology. The new approach ('best estimate' methodology) uses the MCNP5, ORIGEN-2 and BERYL codes to obtain steady-state and decay power distributions for the BR2 core during the tests A/400/1, C/600/3 and F/400/1. This methodology can be easily extended to simulate any BR2 core configuration. Comparisons with measured peak cladding temperatures showed a much better agreement when power distributions obtained with the new methodology are used.

Influence of transient flow on the mobility of strontium in unsaturated sand column

International Nuclear Information System (INIS)

Mazet, P.

2008-10-01

The reactive transport of 85 Sr was studied on laboratory columns, focusing on the influence of transient unsaturated flow (cycles of infiltration and redistribution) associated with controlled geochemistry (constant concentrations of major elements and stable strontium in water). An original experimental tool (gamma attenuation system) allows us to follow at the same time the variations of humidity of the soil and the migration of radionuclide, in a non-destroying and definite way. First stage of this study concerned the implementation of the experimental tool to measure transient hydraulic events within the columns of sand. Several experiments of transport of 85 Sr were then performed with different water condition (saturated, unsaturated, permanent and transient flow). Experimental results were simulated using the computer codes HYDRUS-1D (phenomenological approach with partition coefficient K d ) and HYTEC (mechanistic geochemical/transport approach). Confrontation between experience and modelling shows that, for our operating conditions, transfer of 85 Sr can be predicted with an 'operational' approach using: 1) simplified geochemical model with partition coefficient K d concerning interactive reaction with the soil (K d value determined independently on saturated column, with the same water geochemistry), 2) permanent saturated (or unsaturated) flow, taking into account the cumulated infiltrated water during unsaturated transient hydraulic events concerning hydrodynamic. Generalization of these results (area of validity) suggests that the 'cumulated infiltrated water + K d ' approach can be use, for controlled water geochemistry, when the numerical value of K d is fairly strong (K d ≥≥1), and that it is insensitive to the value of the water content. Moreover, the presence of immobile water (∼10%) recorded with tritium transport, is undetectable with strontium. Explanation of this result is allocated to the different characteristic time residence

Preliminary Assessment of the Loss of Flow Accident for PGSFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Choi, Yong Won; Bae, Moohoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-10-15

TRACE code have being considered as a candidate tool for SFR audit calculation for licensing review since 2012. On the basis of modeling and precalculation experience for the Demonstration Sodium cooled Fast Reactor (DSFR-600), TRACE code model for PGSFR was developed this year. In this paper, one of representing Design Base Event (DBE), Loss of Flow (LOF) accident was pre-calculated and Locked Rotor (LR) case was compared with LOF case since it could be a possible limiting case for LOF representing DBE. Sensitivity calculation for the LR case was implemented for identifying major parameters for the scenario. For the preparation of the review of licensing application for PGSFR, TRACE model for the PGSFR was developed and the loss of flow accident was precalculated. The locked pump rotor case was also calculated as a possible bounding case for the loss of flow scenario. Pre-calculation showed that the locked rotor case was similar or worst case to the loss of flow accident. Therefore, the locked rotor case should take into account in design base accident assessment of PGSFR. Sensitivity calculations for the rocked rotor case also studied for identification of unfixed design parameters influencing to estimation of inner surface temperature. Sensitivity result showed that the first temperature peak was largely influenced by reactor trip delay and second peak mostly influenced by pump coast down characteristic.

4. Workshop - Measurement techniques of stationary and transient multiphase flow

International Nuclear Information System (INIS)

Prasser, H.M.

2001-05-01

In November 2000, the 4th Workshop on Measurement Techniques for Stationary and Transient Multiphase Flows took place in Rossendorf. Three previous workshops of this series were national meetings; this time participants from different countries took part. The programme comprised 14 oral presentations, 9 of which are included in these proceedings in full length. A special highlight of the meeting was the main lecture ''Ultrasonic doppler method for bubbly flow measurement'' of Professor Masanori Aritomi, Dr. Hiroshige Kikura and Dr. Yumiko Suzuki. The workshop again dealt with high-resolution phase distribution and phase velocity measurement techniques based on electrical conductivity, ultrasound, laser light and high-speed cinematography. A number of presentations were dedicated to the application of wire-mesh sensors developed by FZR for different applications used by the Technical Universities of Delft and Munich and the Tokyo Institute of Technology. (orig.)

EVENT, Explosive Transients in Flow Networks

International Nuclear Information System (INIS)

Andrae, R.W.; Tang, P.K.; Bolstad, J.W.; Gregory, W.S.

1985-01-01

1 - Description of problem or function: A major concern of the chemical, nuclear, and mining industries is the occurrence of an explosion in one part of a facility and subsequent transmission of explosive effects through the ventilation system. An explosive event can cause performance degradation of the ventilation system or even structural failures. A more serious consequence is the release of hazardous materials to the environment if vital protective devices such as air filters, are damaged. EVENT was developed to investigate the effects of explosive transients through fluid-flow networks. Using the principles of fluid mechanics and thermodynamics, governing equations for the conservation of mass, energy, and momentum are formulated. These equations are applied to the complete network subdivided into two general components: nodes and branches. The nodes represent boundaries and internal junctions where the conservation of mass and energy applies. The branches can be ducts, valves, blowers, or filters. Since in EVENT the effect of the explosion, not the characteristics of the explosion itself, is of interest, the transient is simulated in the simplest possible way. A rapid addition of mass and energy to the system at certain locations is used. This representation is adequate for all of the network except the region where the explosion actually occurs. EVENT84 is a modification of EVENT which includes a new explosion chamber model subroutine based on the NOL BLAST program developed at the Naval Ordnance Laboratory, Silver Spring, Maryland. This subroutine calculates the confined explosion near-field parameters and supplies the time functions of energy and mass injection. Solid-phase or TNT-equivalent explosions (which simulate 'point source' explosions in nuclear facilities) as well as explosions in gas-air mixtures can be simulated. The four types of explosions EVENT84 simulates are TNT, hydrogen in air, acetylene in air, and tributyl phosphate (TBP or 'red oil

Transient analysis of multifailure conditions by using PWR plant simulator

International Nuclear Information System (INIS)

Morisaki, Hidetoshi; Yokobayashi, Masao.

1984-11-01

This report describes results of the analysis of abnormal transients caused by multifailures using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at the end of life. Various malfunctions to simulate abnormal conditions caused by equipment failures are provided. In this report, features of abnormal transients caused by concurrence of malfunctions are discussed. The abnormal conditions studied are leak of primary coolant, loss of charging and feedwater flows, and control systems failure. From the results, it was observed that transient responses caused by some of the malfunctions are almost same as the addition of behaviors caused by each single malfunction. Therefore, it can be said that kinds of malfunctions which are concurrent may be estimated from transient characteristics of each single malfunction. (author)

FFTF fuel pin design procedure verification for transient operation

International Nuclear Information System (INIS)

Baars, R.E.

1975-05-01

The FFTF design procedures for evaluating fuel pin transient performance are briefly reviewed, and data where available are compared with design procedure predictions. Specifically, burst conditions derived from Fuel Cladding Transient Tester (FCTT) tests and from ANL loss-of-flow tests are compared with burst pressures computed using the design procedure upon which the cladding integrity limit was based. Failure times are predicted using the design procedure for evaluation of rapid reactivity insertion accidents, for five unterminated TREAT experiments in which well characterized fuel failures were deliberately incurred. (U.S.)

A parallel finite-volume finite-element method for transient compressible turbulent flows with heat transfer

International Nuclear Information System (INIS)

Masoud Ziaei-Rad

2010-01-01

In this paper, a two-dimensional numerical scheme is presented for the simulation of turbulent, viscous, transient compressible flows in the simultaneously developing hydraulic and thermal boundary layer region. The numerical procedure is a finite-volume-based finite-element method applied to unstructured grids. This combination together with a new method applied for the boundary conditions allows for accurate computation of the variables in the entrance region and for a wide range of flow fields from subsonic to transonic. The Roe-Riemann solver is used for the convective terms, whereas the standard Galerkin technique is applied for the viscous terms. A modified κ-ε model with a two-layer equation for the near-wall region combined with a compressibility correction is used to predict the turbulent viscosity. Parallel processing is also employed to divide the computational domain among the different processors to reduce the computational time. The method is applied to some test cases in order to verify the numerical accuracy. The results show significant differences between incompressible and compressible flows in the friction coefficient, Nusselt number, shear stress and the ratio of the compressible turbulent viscosity to the molecular viscosity along the developing region. A transient flow generated after an accidental rupture in a pipeline was also studied as a test case. The results show that the present numerical scheme is stable, accurate and efficient enough to solve the problem of transient wall-bounded flow.