Experiment data report for LOFT anticipated transient without scram Experiment L9-4

International Nuclear Information System (INIS)

Batt, D.L.; Divine, J.M.; McKenna, K.J.

1982-11-01

Selected pertinent and uninterpreted data from the fourth anticipated transient with multiple failures experiment (Experiment L9-4) conducted on September 24, 1982, in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system's thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)], commercial PWR operations. Experiment L9-4 simulated a loss-of-offsite-power anticipated transient without reactor scram. The loss-of-offsite-power accident led to an increase in the primary coolant system temperature and pressure. The experiment safety relief valve opened and was able to limit and control the pressure transient. In addition, subsequent heat generation was dissipated by the auxiliary feedwater flow in the secondary coolant system until the reactor was scrammed at experiment termination

Reactive power generation in high speed induction machines by continuously occurring space-transients

Science.gov (United States)

Laithwaite, E. R.; Kuznetsov, S. B.

1980-09-01

A new technique of continuously generating reactive power from the stator of a brushless induction machine is conceived and tested on a 10-kw linear machine and on 35 and 150 rotary cage motors. An auxiliary magnetic wave traveling at rotor speed is artificially created by the space-transient attributable to the asymmetrical stator winding. At least two distinct windings of different pole-pitch must be incorporated. This rotor wave drifts in and out of phase repeatedly with the stator MMF wave proper and the resulting modulation of the airgap flux is used to generate reactive VA apart from that required for magnetization or leakage flux. The VAR generation effect increases with machine size, and leading power factor operation of the entire machine is viable for large industrial motors and power system induction generators.

Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

International Nuclear Information System (INIS)

Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

2013-01-01

Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

Experiment data report for LOFT anticipated transient-without-scram Experiment L9-3

International Nuclear Information System (INIS)

Bayless, P.D.; Divine, J.M.

1982-05-01

Selected pertinent and uninterpreted data from the third anticipated transient with multiple failures experiment (Experiment L9-3) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)], commercial PWR operations. Experiment L9-3 simulated a loss-of-feedwater anticipated transient without scram. The loss-of-feedwater accident led to an increase in the primary coolant system temperature and pressure. Both the experiment power-operated relief valve (PORV) and safety relief valve opened and were able to limit and control the pressure transient. The plant was then recovered with the control rods still withdrawn by injecting 7200-ppM borated water, manually cycling the PORV and feeding and bleeding the steam generator

Study on reactor power transient characteristics (reactor training experiments). Control rod reactivity calibration by positive period method and other experiment

International Nuclear Information System (INIS)

Ozaki, Yoshihiko; Sunagawa, Takeyoshi

2014-01-01

In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity calibration experiments using positive Period method were carried out for shim safety rod and regulating rod, respectively. The results were obtained as reasonable values in comparison with the nominal value of the UTR-KINKI. The measurement of reactor power change after reactor scram was performed, and the presence of the delayed neutron precursor was confirmed by calculating the half-life. The spatial dose rate measurement experiment of neutrons and γ-rays in the reactor room in a reactor power 1W operating conditions were also performed. (author)

Calculations of steady-state and reactivity insertion transients in a research reactor simulating the PWR

International Nuclear Information System (INIS)

Mladin, Mirea; Mladin, Daniela; Prodea, Ilie

2010-01-01

In 2008, IAEA started a Coordinated Research Project for benchmarking the thermalhydraulic and neutronic computer codes for research reactor analysis against the experimental data. In this framework, for the first year of research contract, the Institute for Nuclear Research engaged in steady-state analysis of SPERT-III reactor and also in the simulation of the reactivity insertion tests performed in this reactor during mid sixties. In the first part, the paper describes a Monte Carlo input model of the oxide core selected for investigation and the results of the steady-state neutronic calculations with respect to hot and cold core reactivity excess and control rods worth. Also, prompt neutron life and reactivity feed-back coefficients were examined. These results were compared with the data provided in the reactor specification document concerning neutronic design calculated data. The second part of the paper is dedicated to calculation of the reactivity insertion transients with RELAP5 and CATHARE2 thermalhydraulic codes, both including point reactor kinetics models, and to comparison with experimental data. (authors)

A review of experiments and results from the transient reactor test (TREAT) facility

International Nuclear Information System (INIS)

Deitrich, L. W.

1998-01-01

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

Reactivity estimation using digital nonlinear H∞ estimator for VHTRC experiment

International Nuclear Information System (INIS)

Suzuki, Katsuo; Nabeshima, Kunihiko; Yamane, Tsuyoshi

2003-01-01

On-line and real-time estimation of time-varying reactivity in a nuclear reactor in necessary for early detection of reactivity anomaly and safe operation. Using a digital nonlinear H ∞ estimator, an experiment of real-time dynamic reactivity estimation was carried out in the Very High Temperature Reactor Critical Assembly (VHTRC) of Japan Atomic Energy Research Institute. Some technical issues of the experiment are described, such as reactivity insertion, data sampling frequency, anti-aliasing filter, experimental circuit and digitalising nonlinear H ∞ reactivity estimator, and so on. Then, we discussed the experimental results obtained by the digital nonlinear H ∞ estimator with sampled data of the nuclear instrumentation signal for the power responses under various reactivity insertions. Good performances of estimated reactivity were observed, with almost no delay to the true reactivity and sufficient accuracy between 0.05 cent and 0.1 cent. The experiment shows that real-time reactivity for data sampling period of 10 ms can be certainly realized. From the results of the experiment, it is concluded that the digital nonlinear H ∞ reactivity estimator can be applied as on-line real-time reactivity meter for actual nuclear plants. (author)

Cerebral vasomotor reactivity: steady-state versus transient changes in carbon dioxide tension.

Science.gov (United States)

Brothers, R Matthew; Lucas, Rebekah A I; Zhu, Yong-Sheng; Crandall, Craig G; Zhang, Rong

2014-11-01

Cerebral vasomotor reactivity (CVMR) to changes in arterial carbon dioxide tension (P aCO 2) is assessed during steady-state or transient changes in P aCO 2. This study tested the following two hypotheses: (i) that CVMR during steady-state changes differs from that during transient changes in P aCO 2; and (ii) that CVMR during rebreathing-induced hypercapnia would be blunted when preceded by a period of hyperventilation. For each hypothesis, end-tidal carbon dioxide tension (P ET , CO 2) middle cerebral artery blood velocity (CBFV), cerebrovascular conductance index (CVCI; CBFV/mean arterial pressure) and CVMR (slope of the linear regression between changes in CBFV and CVCI versus P ET , CO 2) were assessed in eight individuals. To address the first hypothesis, measurements were made during the following two conditions (randomized): (i) steady-state increases in P ET , CO 2 of 5 and 10 Torr above baseline; and (ii) rebreathing-induced transient breath-by-breath increases in P ET , CO 2. The linear regression for CBFV versus P ET , CO 2 (P = 0.65) and CVCI versus P ET , CO 2 (P = 0.44) was similar between methods; however, individual variability in CBFV or CVCI responses existed among subjects. To address the second hypothesis, the same measurements were made during the following two conditions (randomized): (i) immediately following a brief period of hypocapnia induced by hyperventilation for 1 min followed by rebreathing; and (ii) during rebreathing only. The slope of the linear regression for CBFV versus P ET , CO 2 (P < 0.01) and CVCI versus P ET , CO 2 (P < 0.01) was reduced during hyperventilation plus rebreathing relative to rebreathing only. These results indicate that cerebral vasomotor reactivity to changes in P aCO 2 is similar regardless of the employed methodology to induce changes in P aCO 2 and that hyperventilation-induced hypocapnia attenuates the cerebral vasodilatory responses during a subsequent period of rebreathing

Modelling Reactivity-Initiated-Accident Experiments With Falcon And SCANAIR: A Comparison Exercise

International Nuclear Information System (INIS)

Romano, A.; Wallin, H.; Zimmermann, M.A.

2005-01-01

A critical assessment is made of the state-of-the-art fuel performance code FALCON in the context of selected Reactivity Initiated Accident (RIA) experiments from the CABRI REP Na series, and contrasts its predictions against those of the extensively benchmarked SCANAIR (Version 3.2) code. The thermal fields in the fuel and cladding, the clad mechanical deformation, and the Fission Gas Release (FGR) are adopted as 'Figures of Merit' by which to judge code performance. Particular attention is paid to the importance of fission-gas-induced clad deformation (which is modelled in SCANAIR, but not in FALCON), relative to that driven by the fuel thermal expansion (which is modelled by both codes). The thermal fields calculated by the codes are in good agreement with each other, especially during the initial stages of the transients --- the adiabatic phase. Larger discrepancies are observed at later times, and are due to the different models applied to calculate the gap conductance. FALCON predicts clad permanent deformations at the end of the transients with a maximum deviation from the experimental measurements of about 20%. Generally, the code always tends to underpredict the measurements. SCANAIR performs similarly, but grossly overpredicts the permanent clad strain for the case involving a very energetic pulse. The fission-gas-driven clad deformation is only relevant for very fast pulse energy injection cases, which are not prototypical of the RIA transients expected in PWRs. The FGR models in FALCON do not capture the mechanism of 'burst-release' in the RIA transients, having been developed for steady-state irradiation conditions. This also explains why they performed poorly when applied to the fast-transient cases analyzed here. In contrast, the FGR results from SCANAIR are in satisfactory agreement with the experimental results. (author)

A FRAMEWORK FOR INTERPRETING FAST RADIO TRANSIENTS SEARCH EXPERIMENTS: APPLICATION TO THE V-FASTR EXPERIMENT

International Nuclear Information System (INIS)

Trott, Cathryn M.; Tingay, Steven J.; Wayth, Randall B.; Macquart, Jean-Pierre R.; Palaniswamy, Divya; Thompson, David R.; Wagstaff, Kiri L.; Majid, Walid A.; Burke-Spolaor, Sarah; Deller, Adam T.; Brisken, Walter F.

2013-01-01

We define a framework for determining constraints on the detection rate of fast transient events from a population of underlying sources, with a view to incorporate beam shape, frequency effects, scattering effects, and detection efficiency into the metric. We then demonstrate a method for combining independent data sets into a single event rate constraint diagram, using a probabilistic approach to the limits on parameter space. We apply this new framework to present the latest results from the V-FASTR experiment, a commensal fast transients search using the Very Long Baseline Array (VLBA). In the 20 cm band, V-FASTR now has the ability to probe the regions of parameter space of importance for the observed Lorimer and Keane fast radio transient candidates by combining the information from observations with differing bandwidths, and properly accounting for the source dispersion measure, VLBA antenna beam shape, experiment time sampling, and stochastic nature of events. We then apply the framework to combine the results of the V-FASTR and Allen Telescope Array Fly's Eye experiments, demonstrating their complementarity. Expectations for fast transients experiments for the SKA Phase I dish array are then computed, and the impact of large differential bandwidths is discussed.

Searching for MHz Transients with the VLA Low-band Ionosphere and Transient Experiment (VLITE)

Science.gov (United States)

Polisensky, Emil; Peters, Wendy; Giacintucci, Simona; Clarke, Tracy; Kassim, Namir E.; hyman, Scott D.; van der Horst, Alexander; Linford, Justin; Waldron, Zach; Frail, Dale

2018-01-01

NRL and NRAO have expanded the low frequency capabilities of the VLA through the VLA Low-band Ionosphere and Transient Experiment (VLITE, http://vlite.nrao.edu/ ), effectively making the instrument two telescopes in one. VLITE is a commensal observing system that harvests data from the prime focus in parallel with normal Cassegrain focus observing on a subset of VLA antennas. VLITE provides over 6000 observing hours per year in a > 5 square degree field-of-view using 64 MHz bandwidth centered on 352 MHz. By operating in parallel, VLITE offers invaluable low frequency data to targeted observations of transient sources detected at higher frequencies. With arcsec resolution and mJy sensitivity, VLITE additionally offers great potential for blind searches of rarer radio-selected transients. We use catalog matching software on the imaging products from the daily astrophysics pipeline and the LOFAR Transients Pipeline (TraP) on repeated observations of the same fields to search for coherent and incoherent astronomical transients on timescales of a few seconds to years. We present the current status of the VLITE transient science program from its initial deployment on 10 antennas in November 2014 through its expansion to 16 antennas in the summer of 2017. Transient limits from VLITE’s first year of operation (Polisensky et al. 2016) are updated per the most recent analysis.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

Energy Technology Data Exchange (ETDEWEB)

Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-05-15

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

International Nuclear Information System (INIS)

Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-01-01

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Modelling of power-reactivity coefficient measurement

International Nuclear Information System (INIS)

Strmensky, C.; Petenyi, V.; Jagrik, J.; Minarcin, M.; Hascik, R.; Toth, L.

2005-01-01

Report describes results of modeling of power-reactivity coefficient analysis on power-level. In paper we calculate values of discrepancies arisen during transient process. These discrepancies can be arisen as result of experiment evaluation and can be caused by disregard of 3D effects on neutron distribution. The results are critically discussed (Authors)

Sensitivity analysis of reactive ecological dynamics.

Science.gov (United States)

Verdy, Ariane; Caswell, Hal

2008-08-01

Ecological systems with asymptotically stable equilibria may exhibit significant transient dynamics following perturbations. In some cases, these transient dynamics include the possibility of excursions away from the equilibrium before the eventual return; systems that exhibit such amplification of perturbations are called reactive. Reactivity is a common property of ecological systems, and the amplification can be large and long-lasting. The transient response of a reactive ecosystem depends on the parameters of the underlying model. To investigate this dependence, we develop sensitivity analyses for indices of transient dynamics (reactivity, the amplification envelope, and the optimal perturbation) in both continuous- and discrete-time models written in matrix form. The sensitivity calculations require expressions, some of them new, for the derivatives of equilibria, eigenvalues, singular values, and singular vectors, obtained using matrix calculus. Sensitivity analysis provides a quantitative framework for investigating the mechanisms leading to transient growth. We apply the methodology to a predator-prey model and a size-structured food web model. The results suggest predator-driven and prey-driven mechanisms for transient amplification resulting from multispecies interactions.

Experiment program and results of the TRACY

International Nuclear Information System (INIS)

Ogawa, K.; Nakajima, K.; Aizawa, E.; Arishima, H.; Morita, T.; Sakuraba, K.; Takahashi, T.; Ohno, A.

1998-01-01

JAERI started supercritical experiments with low enriched uranium nitrate solution with the TRACY in NUCEF. The purpose of the TRACY is to obtain the data on a postulated critical accident phenomena such as power, total number of fissions. In a supercritical experiment, excess reactivity can be inserted by withdrawal of a transient rod or continuous feed of the solution fuel. The TRACY carried out 77 runs including 26 supercritical experiments by the end of 1997. In the transient experiment with reactivity insertion of about 3$, the peak power and the maximum pressure of the core were obtained 1020 MW and 0.50 MPa, respectively. (author)

Using Rising Limb Analysis to Estimate Uptake of Reactive Solutes in Advective and Transient Storage Sub-compartments of Stream Ecosystems

Science.gov (United States)

Thomas, S. A.; Valett, H.; Webster, J. R.; Mulholland, P. J.; Dahm, C. N.

2001-12-01

Identifying the locations and controls governing solute uptake is a recent area of focus in studies of stream biogeochemistry. We introduce a technique, rising limb analysis (RLA), to estimate areal nitrate uptake in the advective and transient storage (TS) zones of streams. RLA is an inverse approach that combines nutrient spiraling and transient storage modeling to calculate total uptake of reactive solutes and the fraction of uptake occurring within the advective sub-compartment of streams. The contribution of the transient storage zones to solute loss is determined by difference. Twelve-hour coinjections of conservative (Cl-) and reactive (15NO3) tracers were conducted seasonally in several headwater streams among which AS/A ranged from 0.01 - 2.0. TS characteristics were determined using an advection-dispersion model modified to include hydrologic exchange with a transient storage compartment. Whole-system uptake was determined by fitting the longitudinal pattern of NO3 to first-order, exponential decay model. Uptake in the advective sub-compartment was determined by collecting a temporal sequence of samples from a single location beginning with the arrival of the solute front and concluding with the onset of plateau conditions (i.e. the rising limb). Across the rising limb, 15NO3:Cl was regressed against the percentage of water that had resided in the transient storage zone (calculated from the TS modeling). The y-intercept thus provides an estimate of the plateau 15NO3:Cl ratio in the absence of NO3 uptake within the transient storage zone. Algebraic expressions were used to calculate the percentage of NO3 uptake occurring in the advective and transient storage sub-compartments. Application of RLA successfully estimated uptake coefficients for NO3 in the subsurface when the physical dimensions of that habitat were substantial (AS/A > 0.2) and when plateau conditions at the sampling location consisted of waters in which at least 25% had resided in the

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.)

Final report for the H3 transient overpower failure threshold experiment

International Nuclear Information System (INIS)

Wright, A.E.; Rothman, A.B.; Stahl, D.; Agrawal, A.K.; Deitrich, L.W.; Chen, S.S.

1975-06-01

Test H3 was the first transient overpower failure-threshold experiment in TREAT to use irradiated fuel and to employ a TREAT transient shaped to produce sample-temperature distributions typical of steady state prior to the overpower excursion. A seven-pin assembly was tested within the Mark-II TREAT sodium loop. The experiment was performed successfully with satisfactory TREAT power transient and loop performance, and demonstrated the capability of intermediate-power EBR-II-irradiated fuel at low burnup (no central void) to withstand a mild overpower transient that terminated with fuel temperatures just short of the solidus without cladding strain. Calculated values of outlet coolant temperature and amount of molten fuel generally agree well with the test data. Posttest thermalhydraulic and mechanical analyses of the fuel pins are reported. Results of detailed nondestructive and destructive examinations of the preirradiated central pin and a fresh peripheral pin are presented. (U.S.)

The reactive solid-gas flow of a fluidized bed for UO2 conversion

International Nuclear Information System (INIS)

Juanico, L.E.

1991-01-01

The reactive solid-gas flow of a fluidized bed for UO 2 conversion was modeled. The sedimentation-reaction process was treated using the drift-flux equations. Also, the associated pressure transient due to the reaction gas release was analyzed. An experiment was carried out to compare the results, and pressure transient was numerically simulated, reaching interesting conclusions. (Author) [es

LMFBR. Off normal, transient test facilities and programs in the USA

International Nuclear Information System (INIS)

Herbst, R.J.

1985-01-01

The United States fast breeder reactor development program has included operational transient analyses and experiments to verify the predicted performance of core components. Operational transient testing has focused on off-normal operation during Plant Protection System terminated transient-overpower events. In-pile and out-of-pile tests have been used to simulate predicted thermal and mechanical strain cycles and measure component response. The spectrum of reactivity ramp rates investigated in TOP tests has recently been expanded to include rates of less than $0.1/s. These slow ramp rate studies are being done in cooperation with the Japanese. The US has also cooperated with the UK in the transient testing of Prototype Fast Reactor fuel pins

Analyses of criticality and reactivity for TRACY experiments based on JENDL-3.3 data library

International Nuclear Information System (INIS)

Sono, Hiroki; Miyoshi, Yoshinori; Nakajima, Ken

2003-01-01

The parameters on criticality and reactivity employed for computational simulations of the TRACY supercritical experiments were analyzed using a recently revised nuclear data library, JENDL-3.3. The parameters based on the JENDL-3.3 library were compared to those based on two former-used libraries, JENDL-3.2 and ENDF/B-VI. In the analyses computational codes, MVP, MCNP version 4C and TWOTRAN, were used. The following conclusions were obtained from the analyses: (1) The computational biases of the effective neutron multiplication factor attributable to the nuclear data libraries and to the computational codes do not depend the TRACY experimental conditions such as fuel conditions. (2) The fractional discrepancies in the kinetic parameters and coefficients of reactivity are within ∼5% between the three libraries. By comparison between calculations and measurements of the parameters, the JENDL-3.3 library is expected to give closer values to the measurements than the JENDL-3.2 and ENDF/B-VI libraries. (3) While the reactivity worth of transient rods expressed in the $ unit shows ∼5% discrepancy between the three libraries according to their respective β eff values, there is little discrepancy in that expressed in the Δk/k unit. (author)

Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

Energy Technology Data Exchange (ETDEWEB)

Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

2009-08-15

The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

What makes ecological systems reactive?

Science.gov (United States)

Snyder, Robin E

2010-06-01

Although perturbations from a stable equilibrium must ultimately vanish, they can grow initially, and the maximum initial growth rate is called reactivity. Reactivity thus identifies systems that may undergo transient population surges or drops in response to perturbations; however, we lack biological and mathematical intuition about what makes a system reactive. This paper presents upper and lower bounds on reactivity for an arbitrary linearized model, explores their strictness, and discusses their biological implications. I find that less stable systems (i.e. systems with long transients) have a smaller possible range of reactivities for which no perturbations grow. Systems with more species have a higher capacity to be reactive, assuming species interactions do not weaken too rapidly as the number of species increases. Finally, I find that in discrete time, reactivity is determined largely by mean interaction strength and neither discrete nor continuous time reactivity are sensitive to food web topology. 2010 Elsevier Inc. All rights reserved.

Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code

International Nuclear Information System (INIS)

Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.

2010-01-01

PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.

Assessment of the TASS 1-D neutronics model for the westinghouse and ABB-CE type PWR reactivity induced transients

International Nuclear Information System (INIS)

Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.

1997-01-01

Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension

Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients

International Nuclear Information System (INIS)

Evaldas, Bubelis; Algirdas, Kaliatka; Eugenijus, Uspuras

2003-01-01

This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500

Experience with transients in German NPPs

International Nuclear Information System (INIS)

Lindauer, E.

1984-01-01

This chapter examines reactor accidents in the Federal Republic of Germany based on the formal reporting system for licensee event reports (LERs) and a special investigation on all unplanned power variations in 3 PWRs. The significant transients experienced by BWR type reactors are analyzed. The main goal is to find weak points which caused the transient or influenced its course in an unfavorable way in order to improve the affected plant and others. The complete survey of all transients, with normally little or no safety relevance, allows statistical evaluations and the analysis of trends. It is concluded that significant transients were mainly experienced at older plants, whereas plants of an advanced design produced very few significant transients. The most frequent human errors which lead to transients are failure search in electronic systems and errors during design and commissioning

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)

Chernobyl reactor transient simulation study

International Nuclear Information System (INIS)

Gaber, F.A.; El Messiry, A.M.

1988-01-01

This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

The assessment of RELAP5/MOD2 based on pressurizer transient experiments

International Nuclear Information System (INIS)

Xue Hanjun; Tanrikut, A.; Menzel, R.

1992-03-01

Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

Periodic transients linked to a variation in reactivity; Transitoires de periode lies a une variation de reactivite

Energy Technology Data Exchange (ETDEWEB)

Furet, J; Weil, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

We study here the influence of the transient, linked to a variation in reactivity, on the measurement of the period, this measurement being made from the logarithmic differential of the power and being defined by 1/T 1/p(dp/dt). We show that the adjustment of the thresholds of period safety is often incompatible with the velocities of liberation of reactivity. A compromise is then necessary between the speed of response of the periodimeter and the speed with which the reactivity is liberated. This makes it necessary to have rapid security devices for the power levels in the piles in which the speeds of liberation of the reactivity are high. (author) [French] On etudie ici l'influence du transitoire lie a une variation de la reactivite sur la mesure de la periode, cette mesure etant faite a partir de la derivee logarithmique de la puissance et etant definie par 1/T 1/p(dp/dt). On montre que le reglage des seuils de securite periode est souvent incompatible avec les vitesses de liberation de reactivite. Il y a alors un compromis a faire entre la vitesse de reponse du periodemetre et la vitesse de liberation de reactivite. Ceci impose de disposer de securites rapides sur les niveaux de puissance, dans les piles ou les vitesses de liberation de reactivite sont importantes. (auteur)

Study on the transient behaviours of MNSR reactor for control rod withdrawal

International Nuclear Information System (INIS)

Yang Shunhai

1995-10-01

The transient behaviours of Miniature Neutron Source Reactor MNSR are analyzed and calculated with the reactor thermohydraulics RETRAN-02 program and the reactor physics MARIA program. The obtained event sequence and consequence from the calculation are compared with the experiments. The effective resonance integral for study on Doppler effect is taken into account. The reactivity temperature coefficient weighting factors are computed. The transient parameters related to reactor power peaking, coolant inlet temperatures, outlet temperatures and coolant mass flow, etc. are computed and compared with the experimental results. (6 refs., 2 figs., 5 tabs.)

Two-frequency, one-detector reactivity system (TFODRS)

International Nuclear Information System (INIS)

Sachs, R.D.; Woodall, D.M.

1985-01-01

A two-frequency, one-detector reactivity system (TFODRS) was experimentally verified on the University of New Mexico (UNM) AGN-201M thermal reactor. That system was used to obtain the absolute steady-state reactivity and to demonstrate the feasibility of acquiring the transient reactivity. A detailed description of TFODRS hardware and software is given in this paper. The TFODRS obtains the absolute and net reactivity by computing the frequency spectrum of the reactor neutron-detector signal. The ratio of the high-frequency to the low-frequency components about an empirical break point is used to determine the reactivity. The TFODRS was successfully used to measure a known AGN-201M steady-state reactivity, with a relative error of 18%. TFODRS transient curves as a function of reactivity were shown to be different from the steady-state curves. The transient curves appear to be a function of the rate of reactivity insertion. The authors speculate that a modified TFODRS, using state-of-the-art microprocessors, could be used for fast reactors. The TFODRS is not presently a practicable reactimeter. However, with more research and development, it is felt it could be used in near-term nuclear industry applications, such as monitoring fuel storage pools

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

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

HEDL W-1 SLSF experiment LOPI transient and boiling test results

International Nuclear Information System (INIS)

Henderson, J.M.; Wood, S.A.; Rothrock, R.B.

1980-01-01

The W-1 Sodium Loop Safety Facility (SLSF) experiment was designed to study the heat release characteristics of fast reactor fuel pins under Loss-of-Piping-Integrity (LOPI) accident conditions and determine stable sodium boiling initiation and recovery limits in a prototypic fuel pin bundle array. The results of the experiment address major second level of assurance (LOA-2) safety issues and provide increased insight and understanding of phenomena that would inherently terminate hypothesized accidents with only limited core damage. The irradiation phase of the experiment, consisting of thirteen individual transients, was performed between May 27 and July 20, 1979. The final transient produced approximately two seconds of coolant boiling, cladding dryout, and incipient fuel pin failure. The facility and test hardware performed as designed, allowing completion of all planned tests and achievement of all test objectives

Transient analysis for resolving safety issues

International Nuclear Information System (INIS)

Chao, J.; Layman, W.

1987-01-01

The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

Transient groundwater chemistry near a river: Effects on U(VI) transport in laboratory column experiments

Science.gov (United States)

Yin, Jun; Haggerty, Roy; Stoliker, Deborah L.; Kent, Douglas B.; Istok, Jonathan D.; Greskowiak, Janek; Zachara, John M.

2011-01-01

In the 300 Area of a U(VI)-contaminated aquifer at Hanford, Washington, USA, inorganic carbon and major cations, which have large impacts on U(VI) transport, change on an hourly and seasonal basis near the Columbia River. Batch and column experiments were conducted to investigate the factors controlling U(VI) adsorption/desorption by changing chemical conditions over time. Low alkalinity and low Ca concentrations (Columbia River water) enhanced adsorption and reduced aqueous concentrations. Conversely, high alkalinity and high Ca concentrations (Hanford groundwater) reduced adsorption and increased aqueous concentrations of U(VI). An equilibrium surface complexation model calibrated using laboratory batch experiments accounted for the decrease in U(VI) adsorption observed with increasing (bi)carbonate concentrations and other aqueous chemical conditions. In the column experiment, alternating pulses of river and groundwater caused swings in aqueous U(VI) concentration. A multispecies multirate surface complexation reactive transport model simulated most of the major U(VI) changes in two column experiments. The modeling results also indicated that U(VI) transport in the studied sediment could be simulated by using a single kinetic rate without loss of accuracy in the simulations. Moreover, the capability of the model to predict U(VI) transport in Hanford groundwater under transient chemical conditions depends significantly on the knowledge of real-time change of local groundwater chemistry.

Design criteria of integrated reactors based on transients

International Nuclear Information System (INIS)

Zanocco, P.; Gimenez, M.; Delmastro, D.

1999-01-01

A new tendency in integrated reactors conceptual design is to include safety criteria through accident analysis. In this work, the effect of design parameters in a Loss of Heat Sink transient using design maps is analyzed. Particularly, geometry related parameters and reactivity coefficients are studied. Also the effect of primary relief/safety valve during the transient is evaluated. A design map for valve area vs. coolant density reactivity coefficient is obtained. A computer code (HUARPE) is developed in order to simulate these transients. Coolant, steam dome, pressure vessel structures and core models are implemented. This code is checked against TRAC with satisfactory results. (author)

Fringe-controlled biodegradation under dynamic conditions: Quasi 2-D flow-through experiments and reactive-transport modeling

Science.gov (United States)

Eckert, Dominik; Kürzinger, Petra; Bauer, Robert; Griebler, Christian; Cirpka, Olaf A.

2015-01-01

Biodegradation in contaminated aquifers has been shown to be most pronounced at the fringe of contaminant plumes, where mixing of contaminated water and ambient groundwater, containing dissolved electron acceptors, stimulates microbial activity. While physical mixing of contaminant and electron acceptor by transverse dispersion has been shown to be the major bottleneck for biodegradation in steady-state plumes, so far little is known on the effect of flow and transport dynamics (caused, e.g., by a seasonally fluctuating groundwater table) on biodegradation in these systems. Towards this end we performed experiments in quasi-two-dimensional flow-through microcosms on aerobic toluene degradation by Pseudomonas putida F1. Plume dynamics were simulated by vertical alteration of the toluene plume position and experimental results were analyzed by reactive-transport modeling. We found that, even after disappearance of the toluene plume for two weeks, the majority of microorganisms stayed attached to the sediment and regained their full biodegradation potential within two days after reappearance of the toluene plume. Our results underline that besides microbial growth, also maintenance and dormancy are important processes that affect biodegradation performance under transient environmental conditions and therefore deserve increased consideration in future reactive-transport modeling.

Additional 5 kWe thermoelectric system temperature transients

International Nuclear Information System (INIS)

Halfen, F.J.

1972-01-01

Several additional system transients have been calculated for the 5 kW(e) TE system and are reported in this document. They include a startup transient with a reactivity rate of 0.005 cents/sec, several startup accidents, a step reactivity insertion at full power and a loss of electrical load. These data are intended for input to system design analyses and for possible use in the protected accident section of the safety report. (U.S.)

Development of real time visual evaluation system for sodium transient thermohydraulic experiments

International Nuclear Information System (INIS)

Tanigawa, Shingo

1990-01-01

A real time visual evaluation system, the Liquid Metal Visual Evaluation System (LIVES), has been developed for the Plant Dynamics Test Loop facility at O-arai Engineering Center. This facility is designed to provide sodium transient thermohydraulic experimental data not only in a fuel subassembly but also in a plant wide system simulating abnormal or accident conditions in liquid metal fast breeder reactors. Since liquid metal sodium is invisible, measurements to obtain experimental data are mainly conducted by numerous thermo couples installed at various locations in the test sections and the facility. The transient thermohydraulic phenomena are a result of complicated interactions among global and local scale three-dimensional phenomena, and short- and long-time scale phenomena. It is, therefore, difficult to grasp intuitively thermohydraulic behaviors and to observe accurately both temperature distribution and flow condition solely by digital data or various types of analog data in evaluating the experimental results. For effectively conducting sodium transient experiments and for making it possible to observe exactly thermohydraulic phenomena, the real time visualization technique for transient thermohydraulics has been developed using the latest Engineering Work Station. The system makes it possible to observe and compare instantly the experiment and analytical results while experiment or analysis is in progress. The results are shown by not only the time trend curves but also the graphic animations. This paper shows an outline of the system and sample applications of the system. (author)

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

Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

Directory of Open Access Journals (Sweden)

Isabelle Guénot-Delahaie

2018-03-01

Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

International Nuclear Information System (INIS)

Raymond, P.; Caruge, D.; Paik, H.J.

1994-01-01

The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

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

Experiment on transient heat transfer in closed narrow channel

International Nuclear Information System (INIS)

Ochiai, Masaaki

1985-01-01

Heat transfer coefficients and transient pressures in closed narrow channels were obtained experimentally, in order to assess the gap heat transfer models in the computer code WTRLGD which were devised to analyze the internal pressure behavior of waterlogged fuel rods. Gap widths of channels are 0.1--0.5mm to simulate the gap region of waterlogged fuel rods, and test fluids are water (7--89.2 0 C) and Freon-113 (9.2 0 C). The results show that the heater temperature and the pressure measured in the experiments without the DNB occurrence are simulated fairly well by the calculational model of WTRLGD where the heat transfer in a closed narrow channel is evaluated with one-dimensional transient thermal conduction equation and Jens and Lottes' correlation for nucleate boiling. Consequently, it is also suggested that the above equations are available for evaluation of heat flux from fuel to internal water of waterlogged fuel rods. The film boiling heat transfer coefficient was in the same order of that evaluated by Bromley's correlation and the DNB heat flux was smaller than that obtained in quasi-steady experiments with ordinary systems, although the experimental data for them were not enough. (author)

Analysis of transients in the SRP test pile

International Nuclear Information System (INIS)

Church, J.P.

1976-11-01

Analysis of the hypothetical upper limit accident in the Savannah River Test Pile showed that the offsite thyroid dose from fission product release would be -3 of the 10-CFR-100 guideline dose for 95 percent of measured meteorological conditions. Offsite whole body dose would be negligible. The Test Pile was modified to limit the length of test piece that can be charged to the pile. These modifications reduce the potential offsite dose to -5 of the regulatory guidelines. Assessment of Test Pile safety included calculations of transients initiated by a variety of reactivity additions that were either terminated or not terminated by safety systems. Reactivity addition mechanisms considered were abnormally driving control rods out of the pile and charging abnormal test pieces into the pile. The transients were evaluated in the adiabatic approximation in which three-dimensional calculations of static flux shapes and reactivity were superimposed on point reactor kinetics calculations. Negative reactivity feedback effects appropriate for the pile and the temperature dependence of material properties, such as specific heat and thermal conductivity, were included. The results show that, for the worst initiators, safety systems can prevent the temperature rise from exceeding 1 0 C anywhere in the Test Pile. If the safety systems do not function, the pile temperatures will increase until the transient is ended by the inherent negative reactivity effects, including the melting of some fuel

Fringe-controlled biodegradation under dynamic conditions: quasi 2-D flow-through experiments and reactive-transport modeling.

Science.gov (United States)

Eckert, Dominik; Kürzinger, Petra; Bauer, Robert; Griebler, Christian; Cirpka, Olaf A

2015-01-01

Biodegradation in contaminated aquifers has been shown to be most pronounced at the fringe of contaminant plumes, where mixing of contaminated water and ambient groundwater, containing dissolved electron acceptors, stimulates microbial activity. While physical mixing of contaminant and electron acceptor by transverse dispersion has been shown to be the major bottleneck for biodegradation in steady-state plumes, so far little is known on the effect of flow and transport dynamics (caused, e.g., by a seasonally fluctuating groundwater table) on biodegradation in these systems. Towards this end we performed experiments in quasi-two-dimensional flow-through microcosms on aerobic toluene degradation by Pseudomonas putida F1. Plume dynamics were simulated by vertical alteration of the toluene plume position and experimental results were analyzed by reactive-transport modeling. We found that, even after disappearance of the toluene plume for two weeks, the majority of microorganisms stayed attached to the sediment and regained their full biodegradation potential within two days after reappearance of the toluene plume. Our results underline that besides microbial growth, also maintenance and dormancy are important processes that affect biodegradation performance under transient environmental conditions and therefore deserve increased consideration in future reactive-transport modeling. Copyright © 2014 Elsevier B.V. All rights reserved.

French experience in transient data collection and fatigue monitoring of PWR's nuclear steam supply system

International Nuclear Information System (INIS)

Sabaton, M.; Morilhat, P.; Savoldelli, D.; Genette, P.

1995-10-01

Electricite de France (EDF), the french national electricity company, is operating 54 standardized pressurizer water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYsteme de Surveillance en FAtigue de la Chaudiere) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After the first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF's nuclear 900 MW is planned after one more year of feedback experience. (authors). 12 refs., 3 figs

A technique for computing bowing reactivity feedback in LMFBR's

International Nuclear Information System (INIS)

Finck, P.J.

1987-01-01

During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme

Interpreting signals from astrophysical transient experiments.

Science.gov (United States)

O'Brien, Paul T; Smartt, Stephen J

2013-06-13

Time-domain astronomy has come of age with astronomers now able to monitor the sky at high cadence, both across the electromagnetic spectrum and using neutrinos and gravitational waves. The advent of new observing facilities permits new science, but the ever-increasing throughput of facilities demands efficient communication of coincident detections and better subsequent coordination among the scientific community so as to turn detections into scientific discoveries. To discuss the revolution occurring in our ability to monitor the Universe and the challenges it brings, on 25-26 April 2012, a group of scientists from observational and theoretical teams studying transients met with representatives of the major international transient observing facilities at the Kavli Royal Society International Centre, UK. This immediately followed the Royal Society Discussion Meeting 'New windows on transients across the Universe' held in London. Here, we present a summary of the Kavli meeting at which the participants discussed the science goals common to the transient astronomy community and analysed how to better meet the challenges ahead as ever more powerful observational facilities come on stream.

Reactivity worth measurement of the control blades of the University of Florida training reactor

International Nuclear Information System (INIS)

Quintero-Leyva, Barbaro

1997-01-01

A series of experiments were carried out in order to measure the reactivity worth of the safety and regulating blades of the University of Florida Training Reactor (UFTR) using the Inverse Kinetics, the Inverse Kinetics-Rod Drop method and the Power Ratio. The reactor's own instrumentation (compensated ion chamber) and an independent counting system (fission chamber) were used. A very smooth exponential decay of the flux was observed after 6s of the beginning of the transients using the reading of the reactor detector. The results of the measurements of the reactivity using both detectors were consistent and in good agreement. The compensated ion chamber showed a very smooth exponential behavior; this suggests that if we could record the power for a small sample time, say 0.1 s from the beginning of the transient, several additional research projects could be accomplished. First, precise intercomparison of the methods could be achieved if the statistics level is acceptable. Second, a precise description of the bouncing of the blades and its effects on the reactivity could be achieved. Finally, the design of a reactivity-meter could be based on such study. (author)

Reactive solute transport in acidic streams

Science.gov (United States)

Broshears, R.E.

1996-01-01

Spatial and temporal profiles of Ph and concentrations of toxic metals in streams affected by acid mine drainage are the result of the interplay of physical and biogeochemical processes. This paper describes a reactive solute transport model that provides a physically and thermodynamically quantitative interpretation of these profiles. The model combines a transport module that includes advection-dispersion and transient storage with a geochemical speciation module based on MINTEQA2. Input to the model includes stream hydrologic properties derived from tracer-dilution experiments, headwater and lateral inflow concentrations analyzed in field samples, and a thermodynamic database. Simulations reproduced the general features of steady-state patterns of observed pH and concentrations of aluminum and sulfate in St. Kevin Gulch, an acid mine drainage stream near Leadville, Colorado. These patterns were altered temporarily by injection of sodium carbonate into the stream. A transient simulation reproduced the observed effects of the base injection.

Summary and evaluation of fuel dynamics transient-overpower experiments: status 1974

International Nuclear Information System (INIS)

Deitrich, L.W.; Doerner, R.C.; Hughes, T.H.; Wright, A.E.

1977-06-01

The report summarizes and evaluates experiments conducted in the Transient Reactor Test Facility (TREAT) using the Mark-II loop facility. The tests discussed are of the E and H series. Detailed descriptions of test conditions and test results as of February 1974 are presented. Since all data have not been acquired on all experiments, this report must be considered interim in nature. Particular emphasis is placed on data relevant to Fast Test Reactor (FTR) safety-analysis efforts

Impact of seasonal forcing on reactive ecological systems.

Science.gov (United States)

Vesipa, Riccardo; Ridolfi, Luca

2017-04-21

Our focus is on the short-term dynamics of reactive ecological systems which are stable in the long term. In these systems, perturbations can exhibit significant transient amplifications before asymptotically decaying. This peculiar behavior has attracted increasing attention. However, reactive systems have so far been investigated assuming that external environmental characteristics remain constant, although environmental conditions (e.g., temperature, moisture, water availability, etc.) can undergo substantial changes due to seasonal cycles. In order to fill this gap, we propose applying the adjoint non-modal analysis to study the impact of seasonal variations of environmental conditions on reactive systems. This tool allows the transient dynamics of a perturbation affecting non-autonomous ecological systems to be described. To show the potential of this approach, a seasonally forced prey-predator model with a Holling II type functional response is studied as an exemplifying case. We demonstrate that seasonalities can greatly affect the transient dynamics of the system. Copyright © 2017 Elsevier Ltd. All rights reserved.

Test on the reactor with the portable digital reactivity meter for physical experiment

International Nuclear Information System (INIS)

Huang Liyuan

2010-01-01

Test must be performed on the zero power reactor During the development of portable digital reactivity meter for physical experiment, in order to check its measurement function and accuracy. It describes the test facility, test core, test methods, test items and test results. The test results show that the instrument satisfy the requirements of technical specification, and satisfy the reactivity measurement in the physical experiments on reactors. (authors)

Operational experience with reactive power control methods optimized for tokamak power supplies

International Nuclear Information System (INIS)

Sihler, C.; Huart, M.; Kaesemann, C.-P.; Streibl, B.

2003-01-01

The power and energy of the ASDEX Upgrade (AUG) tokamak are provided by two separate 10.5 kV, 110-85 Hz networks based on the flywheel generators EZ3-EZ4 in addition to the generator EZ2 dedicated to the toroidal field coil. The 10.5 kV networks supply the thyristor converters allowing fast control of the DC currents in the AUG poloidal field coils. Two methods for improving the load power factor in the present experimental campaign of AUG have been investigated, namely the control of the phase-to-neutral voltage in thyristor converters fitted with neutral thyristors, such as the new 145 MVA modular thyristor converter system (Group 6), and reactive power control achieved by means of static VAr compensators (SVC). The paper shows that reliable compensation up to 90 MVAr was regularly achieved and that electrical transients in SVC modules can be kept at an acceptable level. The paper will discuss the results from the reactive power reduction by SVC and neutral thyristor control and draw a comparative conclusion

Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

International Nuclear Information System (INIS)

Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

1999-02-01

Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

Evaluation of reactivity and Xe behavior during daily load following operation

International Nuclear Information System (INIS)

Sakamoto, Yasunori; Araki, Tsuneyasu; Yamamoto, Fumiaki

1992-01-01

A boiling water reactor (BWR) has an excellent load following capability provided by a core flow control, which is used for changing a reactor power level and for compensating the subsequent Xe concentration change. The core characteristics during load following operations are investigated in detail, using our reactor core simulator. Comparisons of changes of the Doppler reactivity, the void reactivity and the Xe reactivity during transients are performed. Also the features of Xe transient during load following operations are shown. It has been shown that the core flow change required to compensate the Xe reactivity change produces much greater change of the void reactivity than that required for power level changes, and that the resulting local power change in the lower part of the core is greater than that in the upper part, because the Xe concentration change in the lower part is hardly compensated by the core flow control. Also the effects of power level changes, cycle patterns, and initial concentration of Xe and I on the Xe transient behavior have been investigated. (author)

VHTRC experiment for verification test of H∞ reactivity estimation method

International Nuclear Information System (INIS)

Fujii, Yoshio; Suzuki, Katsuo; Akino, Fujiyoshi; Yamane, Tsuyoshi; Fujisaki, Shingo; Takeuchi, Motoyoshi; Ono, Toshihiko

1996-02-01

This experiment was performed at VHTRC to acquire the data for verifying the H∞ reactivity estimation method. In this report, the experimental method, the measuring circuits and data processing softwares are described in details. (author)

KIVA3, Transient Multicomponent 2-D and 3-D Reactive Flows with Fuel Sprays

International Nuclear Information System (INIS)

Amsden, A.A.

2001-01-01

1 - Description of program or function: KIVA3VRELEASE2 is a computer program for the numerical calculation of transient, two and three-dimensional, chemically reactive flows with sprays. It is a newer version of the earlier KIVA3 (1993) that has now been extended to model vertical of canted valves in the cylinder head of a gasoline or diesel engine. KIVA3, in turn, was based on the earlier KIVA2 (1989) and uses the same numerical solution procedure and solves the same sort of equations. KIVA3VRELEASE2 uses a block-structured mesh with connectivity defined through indirect addressing. The departure from a single rectangular structure in logical space allows complex geometries to be modeled with significantly greater efficiency because large regions of deactivated cells are no longer necessary. Cell-face boundary conditions permit greater flexibility and simplification in the application of boundary conditions. KIVA3VRELEASE2 contains a number of significant changes. New features enhance the robustness, efficiency, and usefulness of the overall program for engine modeling. Automatic restart of the cycle with a reduced time-step in case of iteration limit or temperature overflow will reduce code crashes. A new option provides automatic deactivation of a port region when it is closed from the cylinder and reactivation when it communicates with the cylinder. Corrections in the code improve accuracy; extensions to the particle-based liquid wall film model makes the model more complete and a spli injection option has been added. A new subroutine monitors the liquid and gaseous fuel phases and energy balance data and emissions are monitored and printed. New features have been added to the grid generator K3PREP and the graphics post processor, K3POST. 2 - Method of solution: KIVA3VRELEASE2 solves the unsteady equations of motion of a turbulent, chemically reactive mixture of ideal gases, coupled to the equations for a single-component vaporizing fuel spray. The gas

Results from transient transport experiments in Rijnhuizen tokamak project: Heat convection, transport barriers and 'non-local' effects

International Nuclear Information System (INIS)

Mantica, P.; Gorini, G.; Hogeweij, G.M.D.; Kloe, J. de; Lopez Cardozo, N.J.; Schilham, A.M.R.

2001-01-01

An overview of experimental transport studies performed on the Rijnhuizen Tokamak Project (RTP) using transient transport techniques in both Ohmic and ECH dominated plasmas is presented. Modulated Electron Cyclotron Heating (ECH) and oblique pellet injection (OPI) have been used to induce electron temperature (T e ) perturbations at different radial locations. These were used to probe the electron transport barriers observed near low order rational magnetic surfaces in ECH dominated steady-state RTP plasmas. Layers of inward electron heat convection in off-axis ECH plasmas were detected with modulated ECH. This suggests that RTP electron transport barriers consist of heat pinch layers rather than layers of low thermal diffusivity. In a different set of experiments, OPI triggered a transient rise of the core T e due to an increase of the T e gradient in the 1< q<2 region. These transient transport barriers were probed with modulated ECH and found to be due to a transient drop of the electron heat diffusivity, except for off-axis ECH plasmas, where a transient inward pinch is also observed. Transient transport studies in RTP could not solve this puzzling interplay between heat diffusion and convection in determining an electron transport barrier. They nevertheless provided challenging experimental evidence both for theoretical modelling and for future experiments. (author)

Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

International Nuclear Information System (INIS)

Henderson, J.M.; Wood, S.A.; Knight, D.D.

1981-01-01

The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

RETRANS, Reactivity Transients in LWR

International Nuclear Information System (INIS)

Kamelander, G.

1989-01-01

1 - Description of program or function: RETRANS is appropriate to calculate power excursions in light water reactors initiated by reactivity insertions due to withdrawal of control elements. As in the code TWIGL, the neutron physics model is based on the time-dependent two-group neutron diffusion equations. The equation of state of the coolant is approximated by a table built into the code. RETRANS solves the heat conduction equation and calculates the heat transfer coefficient for representative fuel rods at each time-step. 2 - Method of solution: The time-dependent neutron diffusion equations are modified by an exponential transformation and solved by means of a finite difference method. There is an option accelerating the inner iterations of the difference scheme by a coarse-mesh re-balancing method. The heat balance equations of the thermo- hydraulic model are discretized and converted into a tri-diagonal system of linear equations which is solved recursively. 3 - Restrictions on the complexity of the problem: r-z-geometry, one- phase-flow

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

Fast Flux Test Facility (FFTF) feedback reactivity components

International Nuclear Information System (INIS)

Nguyen, D.H.

1988-04-01

The static tests conducted during Cycle 8A (1986) of the FFTF have allowed, for the first time, the experimental determination of each of the feedback reactivities caused by the following mechanisms: fuel axial expansion, control rod repositioning, core radial expansion, and subassembly bowing. A semiempirical equation was obtained to describe each of these feedback components that depended only on the relevant reactor temperature (bowing was presented in a tabular form). The Doppler and sodium density reactivities were calculated using existing mechanistic methods. Although they could also be fitted with closed-form equations depending only on temperatures, these equations are not needed in transient analyses using whole core safety computer codes, which use mechanistic methods. The static feedback reactivity model was extended to obtain a dynamic model via the concept of ''time constants.'' Besides being used for transient analyses in the FFTF, these feedback equations constitute a database for the validation and/or calibration of mechanistic feedback reactivity models. 2 refs., 6 tabs

The integrated circuit IC EMP transient state disturbance effect experiment method investigates

International Nuclear Information System (INIS)

Li Xiaowei

2004-01-01

Transient state disturbance characteristic study on the integrated circuit, IC, need from its coupling path outset. Through cable (aerial) coupling, EMP converts to an pulse current voltage and results in the impact to the integrated circuit I/O orifice passing the cable. Aiming at the armament system construction feature, EMP effect to the integrated circuit, IC inside the system is analyzed. The integrated circuit, IC EMP effect experiment current injection method is investigated and a few experiments method is given. (authors)

Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

2015-01-01

Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

Critical experiment tests of bowing and expansion reactivity calculations for LMRS

International Nuclear Information System (INIS)

Schaefer, R.W.

1988-01-01

Experiments done in several LMR-type critical assemblies simulated core axial expansion, core radial expansion and bowing, coolant expansion, and control driveline expansion. For the most part new experimental techniques were developed to do these experiments. Calculations of the experiments basically used design-level methods, except when it was necessary to investigate complexities peculiar to the experiments. It was found that these feedback reactivities generally are overpredicted, but the predictions are within 30% of the experimental values. 14 refs., 2 figs., 4 tabs

Output of the type 4051 and 4061 period meters during startup transients

International Nuclear Information System (INIS)

Cummins, J.D.

1963-05-01

The report describes a digital computer programme for the Ferranti Mercury computer. With this programme startup transients for the recently developed period meters Types 4051 and 4061 may be computed. The reactivity disturbances considered are steps and terminated ramps of reactivity. Due allowance is taken of the variable time constant which is a feature of these period meters. The reactor may be critical or subcritical initially as desired and the initial input time constant of the period meter may be zero or finite. Some representative transients obtained with the programme are presented and discussed. (author)

Physics experiments with the operating reactor

Energy Technology Data Exchange (ETDEWEB)

Cullington, G R; King, D C

1973-09-27

Experimental techniques have been developed and used on Dragon to give consistent information on excess reactivity and shut down margin. The reactivity measurements have been correlated with the theoretical calculations and have led to improvements in the calculations. The methods used and the results obtained are accepted by the Safety Committee as sufficient evidence for compliance with the fuel loading safety rules. Although the reactor was not designed as an experimental facility, flux and dose measurements experiments have been successfully carried out. Mass flow and negative reactivity transient measurements have been carried out. These are valuable for demonstration of the flexibility of the reactor system and for giving confidence in theoretical calculations.

VHTRC experiment for verification test of H{infinity} reactivity estimation method

Energy Technology Data Exchange (ETDEWEB)

Fujii, Yoshio; Suzuki, Katsuo; Akino, Fujiyoshi; Yamane, Tsuyoshi; Fujisaki, Shingo; Takeuchi, Motoyoshi; Ono, Toshihiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1996-02-01

This experiment was performed at VHTRC to acquire the data for verifying the H{infinity} reactivity estimation method. In this report, the experimental method, the measuring circuits and data processing softwares are described in details. (author).

Twenty-five years of transient counting experience in French PWR units

Energy Technology Data Exchange (ETDEWEB)

Barthelet, B. [Electricite de France (EDF DPN), 93 - Saint-Denis (France); Savoldelli, D.; Fritz, R. [Electricite de France (EDF DPN), 93 - Noisy le Grand (France)

2001-07-01

For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

Twenty-five years of transient counting experience in French PWR units

International Nuclear Information System (INIS)

Barthelet, B.; Savoldelli, D.; Fritz, R.

2001-01-01

For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

Void effects on BWR Doppler and void reactivity feedback

International Nuclear Information System (INIS)

Hsiang-Shou Cheng; Diamond, D.J.

1978-01-01

The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

Poison 1 - a programme for calculation of reactivity transients due to fission product poisoning and its application in continuous determination of xenon and samarium poisoning in reactor KS-150

International Nuclear Information System (INIS)

Rana, S.B.

1973-12-01

The report contains a user's description of the 3-dimensional programme POISON 1 for calculating reactivity transients due to fission-product poisoning during changes of reactor power. The chapter dealing with Xe poisoning contains a description of Xe tables, the method of operational determination of Xe poisoning, use of Xe transients for calibrating control rods and means of shutting down the reactor without being overriden by Xe poisoning. Sm poisoning is determined continuously on the basis of the power diagram of reactor operation. In conclusion a possibility of using the programme in a process computer in combination with self-powered detectors as local power sensors is indicated. (author)

Construction of climate change scenarios from transient climate change experiments for the IPCC impacts assessment

International Nuclear Information System (INIS)

Viner, D.; Hulme, M.; Raper, S.C.B.; Jones, P.D.

1994-01-01

This paper outlines the different methods which may be used for the construction of regional climate change scenarios. The main focus of the paper is the construction of global climate change scenarios from climate change experiments carried out using General Circulation Models (GCMS) An introduction to some GCM climate change experiments highlights the difference between model types and experiments (e.g., equilibrium or transient). The latest generation of climate change experiments has been performed using fully coupled ocean-atmosphere GCMS. These allow transient simulations of climate change to be performed with respect to a given greenhouse gas forcing scenario. There are, however, a number of problems with these simulations which pose difficulties for the construction of climate change scenarios for use in climate change impacts assessment. The characteristics of the transient climate change experiments which pose difficulties for the construction of climate change scenarios are discussed. Three examples of these problems are: different climate change experiments use different greenhouse gas concentration scenarios; the 'cold-start' problem makes it difficult to link future projections of climate change to a given calendar year; a drift of the climate is noticeable in the control simulations. In order to construct climate change scenarios for impacts assessment a method has therefore to be employed which addresses these problems. At present the climate modeling and climate change impacts communities are somewhat polarized in their approach to spatial scales. Current GCMs model the climate at resolutions larger than 2.5 x 3.75 degree, while the majority of impacts assessment studies are undertaken at scales below 50km (or 0.5 degree). This paper concludes by addressing the problems in bringing together these two different modeling perspectives by presenting a number of regional climate change scenarios. 35 refs., 8 figs., 2 tabs

Experiment data report for Loft anticipated transient experiments 16-1, 16-2, and 16-3

International Nuclear Information System (INIS)

Batt, D.L.; Carpenter, J.M.

1980-12-01

This report presents uninterpreted experimental data from the second, third, and fourth anticipated transient experiments (Experiments L6-2, L6-1, and L6-3), conducted in the Loss-of-Fluid Test (LOFT) facility. Experiment L6-2 simulated a loss of forced primary coolant flow in a large PWR by tripping power to primary coolant pump motor generator sets, allowing the pumps to coast down under the influence of the flywheel system. Reactor scram initiated on indication of low flow in the primary coolant system (PCS). Experiment L6-1 simulated a loss of steam load in a large PWR by closing the steam flow control valve which reduced heat removal from the secondary coolant system and caused the PCS temperature and pressure to increase until reactor scram initiated on indication on high PCS pressure. Experiment L6-3 simulated an excessive load increase in a large PWR by opening the steam flow control valve at its maximum rate. PCS temperature and pressure decreased, causing the reactor to scram on indication of low PCS pressure. All experiments were complete when the plant was returned to a hot-standby condition

Lifetimes and reactivities of some 1,2-didehydroazepines commonly used in photoaffinity labeling experiments in aqueous solutions.

Science.gov (United States)

Rizk, Mary S; Shi, Xiaofeng; Platz, Matthew S

2006-01-17

The reactive 1,2-didehydroazepine (cyclic ketenimine) intermediates produced upon photolysis of phenyl azide, 3-hydroxyphenyl azide, 3-methoxyphenyl azide, and 3-nitrophenyl azide in water and in HEPES buffer were studied by laser flash photolysis techniques with UV-vis detection of the transient intermediates. The lifetimes of the 1,2-didehydroazepines were obtained along with the absolute rate constants of their reactions with typical amino acids, nucleosides, and other simple reagents present in a biochemical milieu. The nitro substituent greatly accelerates the bimolecular reactions of the cyclic ketenimines, and the 3-methoxy group greatly decelerates the absolute reactivity of 1,2-didehydroazepines. The intermediate produced by photolysis of 3-hydroxyphenyl azide is much more reactive than the intermediate produced by photolysis of 3-methoxyphenyl azide. We propose that the hydroxyl-substituted 1,2-didehydoazepines rapidly (ketenimines undergo hydrolysis. Azepinones react more rapidly with nucleophiles than do methoxy-substituted 1,2-didehydroazepines and are the active species present upon the photolysis of 3-hydroxyphenyl azide in aqueous solution.

The effect of core configuration on temperature coefficient of reactivity in IRR-1

Energy Technology Data Exchange (ETDEWEB)

Bettan, M.; Silverman, I.; Shapira, M.; Nagler, A. [Soreq Nuclear Research Center, Yavne (Israel)

1997-08-01

Experiments designed to measure the effect of coolant moderator temperature on core reactivity in an HEU swimming pool type reactor were performed. The moderator temperature coefficient of reactivity ({alpha}{sub {omega}}) was obtained and found to be different in two core loadings. The measured {alpha}{sub {omega}} of one core loading was {minus}13 pcm/{degrees}C at the temperature range of 23-30{degrees}C. This value of {alpha}{sub {omega}} is comparable to the data published by the IAEA. The {alpha}{sub {omega}} measured in the second core loading was found to be {minus}8 pcm/{degrees}C at the same temperature range. Another phenomenon considered in this study is core behavior during reactivity insertion transient. The results were compared to a core simulation using the Dynamic Simulator for Nuclear Power Plants. It was found that in the second core loading factors other than the moderator temperature influence the core reactivity more than expected. These effects proved to be extremely dependent on core configuration and may in certain core loadings render the reactor`s reactivity coefficient undesirable.

3-D transient eddy current calculations for the FELIX cylinder experiments

International Nuclear Information System (INIS)

Davey, K.R.; Turner, L.R.

1986-12-01

The three-dimensional eddy current transient field problem is formulated first using the U-V method. This method breaks the vector Helmholtz equation into two scalar Helmholtz equations. Null field integral equations and the appropriate boundary conditions are used to set up an identification matrix which is independent of null field point locations. Embedded in the identification matrix are the unknown eigenvalues of the problem representing its impulse response in time. These eigenvalues are found by equating the determinant of the identification matrix to zero. When this initial forcing function is Fourier decomposed into its spatial harmonics, each Fourier component can be associated with a unique eigenvalue by this technique. The true transient solution comes through a convolution of the impulse response so obtained with the particular external field decay governing the problem at hand. The technique is applied to the FELIX cylinder experiments; computed results are compared to data. A pseudoanalytic confirmation of the eigenvalues so obtained is formulated to validate the procedure

Transient Stability Enhancement in Power System Using Static VAR Compensator (SVC

Directory of Open Access Journals (Sweden)

Youssef MOULOUDI

2012-12-01

Full Text Available In this paper, an indirect adaptive fuzzy excitation and static VAR (unit of reactive power, volt-ampere reactive compensator (SVC controller is proposed to enhance transient stability for the power system, which based on input-output linearization technique. A three-bus system, which contains a generator and static VAR compensator (SVC, is considered in this paper, the SVC is located at the midpoint of the transmission lines. Simulation results show that the proposed controller compared with a controller based on tradition linearization technique can enhance the transient stability of the power system under a large sudden fault, which may occur nearly at the generator bus terminal.

Pius, self-protective thermohydraulics transient without safety system intervention

International Nuclear Information System (INIS)

Fredell, J.; Bredolt, V.

1989-01-01

In this paper, the self-protective thermohydraulic feedback of the PIUS reactor system is illustrated by an in-depth discussion of one typical transient. The selected transient is an undetected total loss of feedwater in the complete absence of conventional safety system intervention. The reactor shuts itself down to residual power in two steps. First, the power decreases due to the strongly negative moderator temperature reactivity coefficient, and then a complete shutdown occurs by ingress of cold, highly borated water from the reactor pool. The transient is terminated without any harm to the fuel or paint systems

Point kinetics improvements to evaluate three-dimensional effects in transients calculation

International Nuclear Information System (INIS)

Castellotti, U.

1987-01-01

A calculation method, which considers the flux axial perturbations in the parameters related to the reactivity within a point kinetics model, is described. The method considered uses axial factors of consideration which act on the thermohydraulic variables included in the reactivity calculation. The PUMA three-dimensional code as reference model for the comparisons, is used. The limitations inherent to the reactivity balance of the point models used in the transients calculation, are given. (Author)

The development of high performance numerical simulation code for transient groundwater flow and reactive solute transport problems based on local discontinuous Galerkin method

International Nuclear Information System (INIS)

Suzuki, Shunichi; Motoshima, Takayuki; Naemura, Yumi; Kubo, Shin; Kanie, Shunji

2009-01-01

The authors develop a numerical code based on Local Discontinuous Galerkin Method for transient groundwater flow and reactive solute transport problems in order to make it possible to do three dimensional performance assessment on radioactive waste repositories at the earliest stage possible. Local discontinuous Galerkin Method is one of mixed finite element methods which are more accurate ones than standard finite element methods. In this paper, the developed numerical code is applied to several problems which are provided analytical solutions in order to examine its accuracy and flexibility. The results of the simulations show the new code gives highly accurate numeric solutions. (author)

The influence of the reactivity ramp on the course of the power transient in the MARK 1A core of the SNR 300

International Nuclear Information System (INIS)

Froehlich, R.; Schmuck, P.

1976-01-01

The course of a hypothetic transient overpower accident caused by the onset of a not further specified reactivity ramp accompanied by the simultaneous failure of both shutdown systems must be analyzed in the SNR 300 Mark 1A core licensing procedure. The present study is limited to the discussion of the starting and shutdown phases of such accidents for the fresh core. Depending on the operational state of the reactor, the core geometry is still intact during the starting phase. In the following shutdown phase (core disassembly phase), large-scale mass transfer leads to the nuclear shutdown of the reactor. (orig./AK) [de

Analysis of void reactivity measurements in full MOX BWR physics experiments

International Nuclear Information System (INIS)

Ando, Yoshihira; Yamamoto, Toru; Umano, Takuya

2008-01-01

In the full MOX BWR physics experiments, FUBILA, four 9x9 test assemblies simulating BWR full MOX assemblies were located in the center of the core. Changing the in-channel moderator condition of the four assemblies from 0% void to 40% and 70% void mock-up, void reactivity was measured using Amplified Source Method (ASM) technique in the subcritical cores, in which three fission chambers were located. ASM correction factors necessary to express the consistency of the detector efficiency between measured core configurations were calculated using collision probability cell calculation and 3D-transport core calculation with the nuclear data library, JENDL-3.3. Measured reactivity worth with ASM correction factor was compared with the calculated results obtained through a diffusion, transport and continuous energy Monte Carlo calculation respectively. It was confirmed that the measured void reactivity worth was reproduced well by calculations. (author)

Transient Go: A Mobile App for Transient Astronomy Outreach

Science.gov (United States)

Crichton, D.; Mahabal, A.; Djorgovski, S. G.; Drake, A.; Early, J.; Ivezic, Z.; Jacoby, S.; Kanbur, S.

2016-12-01

Augmented Reality (AR) is set to revolutionize human interaction with the real world as demonstrated by the phenomenal success of `Pokemon Go'. That very technology can be used to rekindle the interest in science at the school level. We are in the process of developing a prototype app based on sky maps that will use AR to introduce different classes of astronomical transients to students as they are discovered i.e. in real-time. This will involve transient streams from surveys such as the Catalina Real-time Transient Survey (CRTS) today and the Large Synoptic Survey Telescope (LSST) in the near future. The transient streams will be combined with archival and latest image cut-outs and other auxiliary data as well as historical and statistical perspectives on each of the transient types being served. Such an app could easily be adapted to work with various NASA missions and NSF projects to enrich the student experience.

Method of analysis to determine subcritical reactivity from the pulsed neutron experiment

International Nuclear Information System (INIS)

Parks, P.B.

1975-06-01

The published methods for the deduction of reactivity from pulsed neutron experiments on subcritical reactors are reviewed. Each method is categorized as inherently yielding a result that is either spatially independent or spatially dependent. The spatially independent results are formally identical with the static reactivity; the result does not depend, in principle, on the location of either the pulsed neutron source or the neutron detector during data collection. The spatially dependent results only approximate the static reactivity; the results are affected, in varying degrees, by the locations of the source and detector. Among the techniques yielding spatially independent results are the Space-Time method of Parks and Stewart and the Inhour method of Preskitt et al. Spatially dependent results are obtained with the Sjoestrand, Gozani, and Garelis-Russell methods which are examined with and without the kinetic distortion corrections given by Becker and Quisenberry. Intercomparisons of all methods are made with reference to pulsed neutron experiments on both unreflected and reflected reactors. Recommendations are made concerning the best choice of method under the various experimental conditions that are likely to be encountered. 14 references. (U.S.)

Reactivity balance for a soluble boron-free small modular reactor

Directory of Open Access Journals (Sweden)

Lezani van der Merwe

2018-06-01

Full Text Available Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR design, only control rods are available to control such rapid core transient.The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model.The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power. Keywords: Control Rod Worth, Reactivity Balance, Reactivity Feedback, Small Modular Reactor, Soluble Boron Free

Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

Energy Technology Data Exchange (ETDEWEB)

Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

2012-07-01

The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)

Classification of transient processes with a jumplike change in the reactivity

International Nuclear Information System (INIS)

Sabaeva, T.A.

1989-01-01

The problem of the change in the neutron flux density accompanying a jumplike (instantaneous) change in the reactivity is classical and is studied in most textbooks and monographs devoted to the regulation of nuclear reactors, where in constructing the response only the feedback on delayed neutrons is taken into account. The use of a linear feedback of a general form permits describing reactors of different types. A classification of feedbacks on reactivity was presented by Sabaeva, where a parabolic region in phase space is separated. A peak in the neutron flux corresponds to the image point falling into this region. In this paper the conditions making it possible to find the change in the neutrons flux immediately after an instantaneous change in the reactivity are derived, and the feedbacks are classified based on this

LIMITS ON THE EVENT RATES OF FAST RADIO TRANSIENTS FROM THE V-FASTR EXPERIMENT

International Nuclear Information System (INIS)

Wayth, Randall B.; Tingay, Steven J.; Deller, Adam T.; Brisken, Walter F.; Thompson, David R.; Wagstaff, Kiri L.; Majid, Walid A.

2012-01-01

We present the first results from the V-FASTR experiment, a commensal search for fast transient radio bursts using the Very Long Baseline Array (VLBA). V-FASTR is unique in that the widely spaced VLBA antennas provide a discriminant against non-astronomical signals and a mechanism for the localization and identification of events that is not possible with single dishes or short baseline interferometers. Thus, far V-FASTR has accumulated over 1300 hr of observation time with the VLBA, between 90 cm and 3 mm wavelength (327 MHz-86 GHz), providing the first limits on fast transient event rates at high radio frequencies (>1.4 GHz). V-FASTR has blindly detected bright individual pulses from seven known pulsars but has not detected any single-pulse events that would indicate high-redshift impulsive bursts of radio emission. At 1.4 GHz, V-FASTR puts limits on fast transient event rates comparable with the PALFA survey at the Arecibo telescope, but generally at lower sensitivities, and comparable to the 'fly's eye' survey at the Allen Telescope Array, but with less sky coverage. We also illustrate the likely performance of the Phase 1 SKA dish array for an incoherent fast transient search fashioned on V-FASTR.

Code Coupling for Multi-Dimensional Core Transient Analysis

International Nuclear Information System (INIS)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

2015-01-01

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

Code Coupling for Multi-Dimensional Core Transient Analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

2015-05-15

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

Update on the Commensal VLA Low-band Ionospheric and Transient Experiment (VLITE)

Science.gov (United States)

Kassim, Namir E.; Clarke, Tracy E.; Ray, Paul S.; Polisensky, Emil; Peters, Wendy M.; Giacintucci, Simona; Helmboldt, Joseph F.; Hyman, Scott D.; Brisken, Walter; Hicks, Brian; Deneva, Julia S.

2017-01-01

The JVLA Low-band Ionospheric and Transient Experiment (VLITE) is a commensal observing system on the NRAO JVLA. The separate optical path of the prime-focus sub-GHz dipole feeds and the Cassegrain-focus GHz feeds provided an opportunity to expand the simultaneous frequency operation of the JVLA through joint observations across both systems. The low-band receivers on 10 JVLA antennas are outfitted with dedicated samplers and use spare fibers to transport the 320-384 MHz band to the VLITE correlator. The initial phase of VLITE uses a custom-designed real-time DiFX software correlator to produce autocorrelations, as well as parallel and cross-hand cross-correlations from the linear dipole feeds. NRL and NRAO have worked together to explore the scientific potential of the commensal low frequency system for ionospheric remote sensing, astrophysics and transients. VLITE operates at nearly 70% wall time with roughly 6200 hours of JVLA time recorded each year.VLITE data are used in real-time for ionospheric research and are transferred daily to NRL for processing in the astrophysics and transient pipelines. These pipelines provide automated radio frequency interference excision, calibration, imaging and self-calibration of data.We will review early scientific results from VLITE across all three science focus areas, including the ionosphere, slow (> 1 sec) transients, and astrophysics. We also discuss the future of the project, that includes its planned expansion to eVLITE including the addition of more antennas, and a parallel capability to search for fast (complement NRAO’s 3 GHz VLA Sky Survey (VLASS). Revised pipelines are under development for operation during the on-the-fly operation mode of the sky survey.

The impact of fuel temperature reactivity coefficient on loss of reactivity control accident

International Nuclear Information System (INIS)

Park, J. H.; Ryu, E. H.; Song, Y. M.; Jung, J. Y.

2012-01-01

Nuclear reactors experience small power fluctuations or anticipated operational transients during even normal power operation. During normal operation, the reactivity is mainly controlled by liquid zone controllers, adjuster rods, mechanical control absorbers, and moderator poison. Even when the reactor power is increased abruptly and largely from an accident and when reactor control systems cannot be actuated quickly due to a fast transient, the reactor should be controlled and stabilized by its inherent safety parameter, such as a negative PCR (Power Coefficient of Reactivity) feedback. A PWR (Pressurized Water Reactor), it is well designed for the reactor to have a negative PCR so that the reactor can be safely shut down or stabilized whenever an abrupt reactivity insertion into the reactor core occurs or the reactor power is abruptly increased. However, it is known that a CANDU reactor has a small amount of PCR, as either negative or positive, because of the different design basis and safety concepts from a PWR. CNSC's regulatory and safety regime has stated that; The PCR of CANDU reactors does not pose a significant risk. Consistent with Canadian nuclear safety requirements, nuclear power plants must have an appropriate combination of inherent and engineered safety features incorporated into the design of the reactor safety and control systems. A reactor design that has a PCR is quite acceptable provided that the reactor is stable against power fluctuations, and that the probability and consequences of any potential accidents that would be aggravated by a positive reactivity feedback are maintained within CNSCprescribed limits. Recently, it was issued licensing the refurbished Wolsong unit 1 in Korea to be operated continuously after its design lifetime in which the calculated PCR was shown to have a small positive value by applying the recent physics code systems, which are composed of WIMS IST, DRAGON IST, and RFSP IST. These code systems were transferred

Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

Science.gov (United States)

Clamens, Olivier; Lecerf, Johann; Hudelot, Jean-Pascal; Duc, Bertrand; Cadiou, Thierry; Blaise, Patrick; Biard, Bruno

2018-01-01

CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN) and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

Directory of Open Access Journals (Sweden)

Clamens Olivier

2018-01-01

Full Text Available CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

Radial core expansion reactivity feedback in advanced LMRs: uncertainties and their effects on inherent safety

International Nuclear Information System (INIS)

Wigeland, R.A.; Moran, T.J.

1988-01-01

An analytical model for calculating radial core expansion, based on the thermal and elastic bowing of a single subassembly at the core periphery, is used to quantify the effect of uncertainties on this reactivity feedback mechanism. This model has been verified and validated with experimental and numerical results. The impact of these uncertainties on the safety margins in unprotected transients is investigated with SASSYS/SAS4A, which includes this model for calculating the reactivity feedback from radial core expansion. The magnitudes of these uncertainties are not sufficient to preclude the use of radial core expansion reactivity feedback in transient analysis

Reactivation of BK polyomavirus in patients with multiple sclerosis receiving natalizumab therapy.

LENUS (Irish Health Repository)

Lonergan, Roisin M

2012-02-01

Natalizumab therapy in multiple sclerosis has been associated with JC polyomavirus-induced progressive multifocal leucoencephalopathy. We hypothesized that natalizumab may also lead to reactivation of BK, a related human polyomavirus capable of causing morbidity in immunosuppressed groups. Patients with relapsing remitting multiple sclerosis treated with natalizumab were prospectively monitored for reactivation of BK virus in blood and urine samples, and for evidence of associated renal dysfunction. In this cohort, JC and BK DNA in blood and urine; cytomegalovirus (CMV) DNA in blood and urine; CD4 and CD8 T-lymphocyte counts and ratios in peripheral blood; and renal function were monitored at regular intervals. BK subtyping and noncoding control region sequencing was performed on samples demonstrating reactivation. Prior to commencement of natalizumab therapy, 3 of 36 patients with multiple sclerosis (8.3%) had BK viruria and BK reactivation occurred in 12 of 54 patients (22.2%). BK viruria was transient in 7, continuous in 2 patients, and persistent viruria was associated with transient viremia. Concomitant JC and CMV viral loads were undetectable. CD4:CD8 ratios fluctuated, but absolute CD4 counts did not fall below normal limits. In four of seven patients with BK virus reactivation, transient reductions in CD4 counts were observed at onset of BK viruria: these resolved in three of four patients on resuppression of BK replication. No renal dysfunction was observed in the cohort. BK virus reactivation can occur during natalizumab therapy; however, the significance in the absence of renal dysfunction is unclear. We propose regular monitoring for BK reactivation or at least for evidence of renal dysfunction in patients receiving natalizumab.

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

Experimental validation of Pu-Sm evolution model for CANDU-6 power transients

International Nuclear Information System (INIS)

Coutsiers, Eduardo E.; Pomerantz, Marcelo E.; Moreno, Carlos A.

2000-01-01

Development of a methodology to evaluate the reactivity produced by Pu-Sm transient, effect displayed after power transients. This methodology allows to predict the behavior of liquid zones with which the fine control of CANDU reactor power is made. With this information, it is easier to foresee the refueling demand after power movements. The comparison with experimental results showed good agreement. (author)

Heat shock and herpes virus: enhanced reactivation without untargeted mutagenesis

International Nuclear Information System (INIS)

Lytle, C.D.; Carney, P.G.

1988-01-01

Enhanced reactivation of Ultraviolet-irradiated virus has been reported to occur in heat-shocked host cells. Since enhanced virus reactivation is often accompanied by untargeted mutagenesis, we investigated whether such mutagenesis would occur for herpes simplex virus (HSV) in CV-1 monkey kidney cells subjected to heat shock. In addition to expressing enhanced reactivation, the treated cells were transiently more susceptible to infection by unirradiated HSV. No mutagenesis of unirradiated HSV was found whether infection occurred at the time of increased susceptibility to infection or during expression of enhanced viral reactivation

Actinide and Xenon reactivity effects in ATW high flux systems

International Nuclear Information System (INIS)

Woosley, M.; Olson, K.; Henderson, D.L.

1995-01-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

Actinide and xenon reactivity effects in ATW high flux systems

International Nuclear Information System (INIS)

Woosley, M.; Olson, K.; Henderson, D. L.; Sailor, W. C.

1995-01-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

Actinide and Xenon reactivity effects in ATW high flux systems

Energy Technology Data Exchange (ETDEWEB)

Woosley, M. [Univ. of Virginia, Charlottesville, VA (United States); Olson, K.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)] [and others

1995-10-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides.

New developments in French transient monitoring system: SYSFAC From the experiments to the industrial process

Energy Technology Data Exchange (ETDEWEB)

Balley, J. [Electricite de France, 69 - Villeurbanne (France). SEPTEN; Bertagnolio, D. [Electricite de France, 69 - Villeurbanne (France). SEPTEN; Faidy, C. [Electricite de France, 69 - Villeurbanne (France). SEPTEN; Kappler, F. [Electricite de France, 69 - Villeurbanne (France). SEPTEN; Kergoat, M. [Electricite de France, 69 - Villeurbanne (France). SEPTEN; L`Huby, Y. [Electricite de France, 69 - Villeurbanne (France). SEPTEN; Genette, P. [Electricite de France, 69 - Villeurbanne (France). SEPTEN; Savoldelli, D. [Electricite de France, Production-Transport DMAINT, 13 Esplanade Charles de Gaulle, 92060 La Defense (France); Fournier, I. [Electricite de France, Direction Etudes et Recherches REME, 25 allee privee, Carrefour Pleyel, 93206 St. Denis (France)

1995-01-01

After more than 15 years of experience with regulatory transient data collection, Electricite de France decided to design a new concept of fatigue monitoring system called SYSFAC. This new system is the result of seven years of successful experimentation with fatigue meters. This system will be connected to the on-site data acquisition system without any complementary instrumentation. The SYSFAC system has a modular structure: the mechanical transient module, the functional transient module, the fatigue meters module and the global damage computing module all have a high level of flexibility to be applied to various types of circuits. After the preliminary studies had been achieved, it was decided to undertake the industrial phase of the SYSFAC project. Specific codes on PC computers have been used to validate the basic concepts and the operator interface. Real-size coding will last one year and the first SYSFAC system will be delivered to the pilot power plant by the end of 1995. ((orig.)).

New developments in French transient monitoring system: SYSFAC From the experiments to the industrial process

International Nuclear Information System (INIS)

Balley, J.; Fournier, I.

1995-01-01

After more than 15 years of experience with regulatory transient data collection, Electricite de France decided to design a new concept of fatigue monitoring system called SYSFAC. This new system is the result of seven years of successful experimentation with fatigue meters. This system will be connected to the on-site data acquisition system without any complementary instrumentation. The SYSFAC system has a modular structure: the mechanical transient module, the functional transient module, the fatigue meters module and the global damage computing module all have a high level of flexibility to be applied to various types of circuits. After the preliminary studies had been achieved, it was decided to undertake the industrial phase of the SYSFAC project. Specific codes on PC computers have been used to validate the basic concepts and the operator interface. Real-size coding will last one year and the first SYSFAC system will be delivered to the pilot power plant by the end of 1995. ((orig.))

Transient analyzer

International Nuclear Information System (INIS)

Muir, M.D.

1975-01-01

The design and design philosophy of a high performance, extremely versatile transient analyzer is described. This sub-system was designed to be controlled through the data acquisition computer system which allows hands off operation. Thus it may be placed on the experiment side of the high voltage safety break between the experimental device and the control room. This analyzer provides control features which are extremely useful for data acquisition from PPPL diagnostics. These include dynamic sample rate changing, which may be intermixed with multiple post trigger operations with variable length blocks using normal, peak to peak or integrate modes. Included in the discussion are general remarks on the advantages of adding intelligence to transient analyzers, a detailed description of the characteristics of the PPPL transient analyzer, a description of the hardware, firmware, control language and operation of the PPPL transient analyzer, and general remarks on future trends in this type of instrumentation both at PPPL and in general

Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments

International Nuclear Information System (INIS)

Sato, Ikken; Lemoine, Francette; Struwe, Dankward

2004-01-01

In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear density and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long

Biogeochemical processes in a clay formation in situ experiment: Part F - Reactive transport modelling

Energy Technology Data Exchange (ETDEWEB)

Tournassat, Christophe, E-mail: c.tournassat@brgm.fr [BRGM, French Geological Survey, Orleans (France); Alt-Epping, Peter [Rock-Water Interaction Group, Institute of Geological Sciences, University of Bern (Switzerland); Gaucher, Eric C. [BRGM, French Geological Survey, Orleans (France); Gimmi, Thomas [Rock-Water Interaction Group, Institute of Geological Sciences, University of Bern (Switzerland)] [Laboratory for Waste Management, Paul Scherrer Institut, Villigen (Switzerland); Leupin, Olivier X. [NAGRA, CH-5430 Wettingen (Switzerland); Wersin, Paul [Gruner Ltd., CH-4020 Basel (Switzerland)

2011-06-15

Highlights: > Reactive transport modelling was used to simulate simultaneously solute transport, thermodynamic reactions, ion exchange and biodegradation during an in-situ experiment in a clay-rock formation. > Opalinus clay formation has a high buffering capacity in terms of chemical perturbations caused by bacterial activity. > Buffering capacity is mainly attributed to the carbonate system and to the reactivity of clay surfaces (cation exchange, pH buffering). - Abstract: Reactive transport modelling was used to simulate solute transport, thermodynamic reactions, ion exchange and biodegradation in the Porewater Chemistry (PC) experiment at the Mont Terri Rock Laboratory. Simulations show that the most important chemical processes controlling the fluid composition within the borehole and the surrounding formation during the experiment are ion exchange, biodegradation and dissolution/precipitation reactions involving pyrite and carbonate minerals. In contrast, thermodynamic mineral dissolution/precipitation reactions involving alumo-silicate minerals have little impact on the fluid composition on the time-scale of the experiment. With the accurate description of the initial chemical condition in the formation in combination with kinetic formulations describing the different stages of bacterial activities, it has been possible to reproduce the evolution of important system parameters, such as the pH, redox potential, total organic C, dissolved inorganic C and SO{sub 4} concentration. Leaching of glycerol from the pH-electrode may be the primary source of organic material that initiated bacterial growth, which caused the chemical perturbation in the borehole. Results from these simulations are consistent with data from the over-coring and demonstrate that the Opalinus Clay has a high buffering capacity in terms of chemical perturbations caused by bacterial activity. This buffering capacity can be attributed to the carbonate system as well as to the reactivity of

Power and power-to-flow reactivity transfer functions in EBR-II [Experimental Breeder Reactor II] fuel

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1989-01-01

Reactivity transfer functions are important in determining the reactivity history during a power transient. Overall nodal transfer functions have been calculated for different subassembly types in the Experimental Breeder Reactor II (EBR-II). Steady-state calculations for temperature changes and, hence, reactivities for power changes have been separated into power and power-to-flow-dependent terms. Axial nodal transfer functions separated into power and power-to-flow-dependent components are reported in this paper for a typical EBR-II fuel pin. This provides an improved understanding of the time dependence of these components in transient situations

Coherent optical transients observed in rubidium atomic line filtered Doppler velocimetry experiments

Energy Technology Data Exchange (ETDEWEB)

Fajardo, Mario E., E-mail: mario.fajardo@eglin.af.mil; Molek, Christopher D.; Vesely, Annamaria L. [Air Force Research Laboratory, Munitions Directorate, Ordnance Division, Energetic Materials Branch, AFRL/RWME, 2306 Perimeter Road, Eglin AFB, Florida 32542-5910 (United States)

2015-10-14

We report the first successful results from our novel Rubidium Atomic Line Filtered (RALF) Doppler velocimetry apparatus, along with unanticipated oscillatory signals due to coherent optical transients generated within pure Rb vapor cells. RALF is a high-velocity and high-acceleration extension of the well-known Doppler Global Velocimetry (DGV) technique for constructing multi-dimensional flow velocity vector maps in aerodynamics experiments [H. Komine, U.S. Patent No. 4,919,536 (24 April 1990)]. RALF exploits the frequency dependence of pressure-broadened Rb atom optical absorptions in a heated Rb/N{sub 2} gas cell to encode the Doppler shift of reflected near-resonant (λ{sub 0} ≈ 780.24 nm) laser light onto the intensity transmitted by the cell. The present RALF apparatus combines fiber optic and free-space components and was built to determine suitable operating conditions and performance parameters for the Rb/N{sub 2} gas cells. It yields single-spot velocities of thin laser-driven-flyer test surfaces and incorporates a simultaneous Photonic Doppler Velocimetry (PDV) channel [Strand et al., Rev. Sci. Instrum. 77, 083108 (2006)] for validation of the RALF results, which we demonstrate here over the v = 0 to 1 km/s range. Both RALF and DGV presume the vapor cells to be simple Beer's Law optical absorbers, so we were quite surprised to observe oscillatory signals in experiments employing low pressure pure Rb vapor cells. We interpret these oscillations as interference between the Doppler shifted reflected light and the Free Induction Decay (FID) coherent optical transient produced within the pure Rb cells at the original laser frequency; this is confirmed by direct comparison of the PDV and FID signals. We attribute the different behaviors of the Rb/N{sub 2} vs. Rb gas cells to efficient dephasing of the atomic/optical coherences by Rb-N{sub 2} collisions. The minimum necessary N{sub 2} buffer gas density ≈0.3 amagat translates into a

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

International Nuclear Information System (INIS)

Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

2015-01-01

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

Energy Technology Data Exchange (ETDEWEB)

Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

2015-02-15

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State

2016-11-30

The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup

Proceedings of the OECD/CSNI specialists meeting on boron dilution reactivity transients

International Nuclear Information System (INIS)

1997-06-01

The purpose of the meeting was to bring together experts involved in the different activities related to boron dilution transients. The experts came from all involved parties, including research organizations, regulatory authorities, vendors and utilities. Information was openly shared and discussed on the experimental results, plant and systems analysis, numerical analysis of mixing and probability and consequences of these transients. Regulatory background and licensing implications were also included to provide the proper frame work for the technical discussion. Each of these areas corresponded to a separate session. The meeting focused on the thermal-hydraulic aspects because of the current interest in that subject and the significant amount of new technical information being generated

Experiments in ZED-2 to study the physics of low-void reactivity fuel in CANDU

International Nuclear Information System (INIS)

Zeller, M.B.; Celli, A.; McPhee, G.P.

1994-01-01

Prospective CANDU clients have indicated a desire for a zero or negative coolant void reactivity. In response to this market requirement AECL Research and AECL CANDU are jointly developing and testing a Low-Void Reactivity Fuel (LVRF) bundle, which will be retrofitable to the current generation of CANDU reactors. An important component of the LVRF program is the undertaking of reactor-physics experiments in the zero-energy ZED-2 lattice test facility at Chalk River Laboratories. Preliminary void-reactivity measurements have already been performed in ZED-2 using a limited amount of the prototype fuel. These experiments were to provide a proof-of-principle for the LVRF concept. A more comprehensive set of experiments are planned for later this year. Experiments to be performed include: measuring the critical buckling of CANDU-type lattices containing LVRF, with and without coolant in the channels; measuring the reactivity effect of heating the LVRF fuel and coolant in ZED-2 hot channels; and measuring detailed reaction rates and neutron density distributions across a LVRF bundle, in voided and D 2 O-cooled channels, by the foil activation method. This paper describes the experimental approach to be used for the study and presents calculations employing transport and diffusion theory to predict the results. The codes used for the simulations are the lattice code WIMS-AECL and the core code CONIFERS. Included in the paper are results from the preliminary measurement of void coefficient for LVRF in a ZED-2 lattice and a comparison of those results to predictions based on WIMS-AECL calculations. (author). 3 refs., 1 tab., 10 figs

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system

International Nuclear Information System (INIS)

Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

1980-05-01

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin

HEDL experimental transient overpower program

International Nuclear Information System (INIS)

Hikido, T.; Culley, G.E.

1976-01-01

HEDL is conducting a series of experiments to evaluate the performance of Fast Flux Test Facility (FFTF) prototypic fuel pins up to the point of cladding breach. A primary objective of the program is to demonstrate the adequacy of fuel pin and Plant Protective System (PPS) designs for terminated transients. Transient tests of prototypic FFTF fuel pins previously irradiated in the Experimental Breeder Reactor-II (EBR-II) have demonstrated the adequacy of the PPS and fuel pin designs and indicate that a very substantial margin exists between PPS-terminated transients and that required to produce fuel pin cladding failure. Additional experiments are planned to extend the data base to high burnup, high fluence fuel pin specimens

Detailed characterization of a Comparative Reactivity Method (CRM) instrument for ambient OH reactivity measurements: experiments vs. modeling

Science.gov (United States)

Michoud, Vincent; Locoge, Nadine; Dusanter, Sébastien

2015-04-01

The Hydroxyl radical (OH) is the main daytime oxidant in the troposphere, leading to the oxidation of Volatile Organic Compounds (VOCs) and the formation of harmful pollutants such as ozone (O3) and Secondary Organic Aerosols (SOA). While OH plays a key role in tropospheric chemistry, recent studies have highlighted that there are still uncertainties associated with the OH budget, i.e the identification of sources and sinks and the quantification of production and loss rates of this radical. It has been demonstrated that ambient measurements of the total OH loss rate (also called total OH reactivity) can be used to identify and reduce these uncertainties. In this context, the Comparative Reactivity Method (CRM), developed by Sinha et al. (ACP, 2008), is a promising technique to measure total OH reactivity in ambient air and has already been used during several field campaigns. This technique relies on monitoring competitive reactions of OH with ambient trace gases and a reference compound (pyrrole) in a sampling reactor to derive ambient OH reactivity. However, this technique requires a complex data processing chain that has yet to be carefully investigated in the laboratory. In this study, we present a detailed characterization of a CRM instrument developed at Mines Douai, France. Experiments have been performed to investigate the dependence of the CRM response on humidity, ambient NOx levels, and the pyrrole-to-OH ratio inside the sampling reactor. Box modelling of the chemistry occurring in the reactor has also been performed to assess our theoretical understanding of the CRM measurement. This work shows that the CRM response is sensitive to both humidity and NOx, which can be accounted for during data processing using parameterizations depending on the pyrrole-to-OH ratio. The agreement observed between laboratory studies and model results suggests a good understanding of the chemistry occurring in the sampling reactor and gives confidence in the CRM

Transients: The regulator's view

International Nuclear Information System (INIS)

Sheron, B.W.; Speis, T.P.

1984-01-01

This chapter attempts to clarify the basis for the regulator's concerns for transient events. Transients are defined as both anticipated operational occurrences and postulated accidents. Recent operational experience, supplemented by improved probabilistic risk analysis methods, has demonstrated that non-LOCA transient events can be significant contributors to overall risk. Topics considered include lessons learned from events and issues, the regulations governing plant transients, multiple failures, different failure frequencies, operator errors, and public pressure. It is concluded that the formation of Owners Groups and Regulatory Response Groups within the owners groups are positive signs of the industry's concern for safety and responsible dealing with the issues affecting both the US NRC and the industry

French experience in transient data collection and fatigue monitoring of PWR`s nuclear steam supply system; Experience francaise sur la comptabilisation des transitoires et la surveillance en fatigue des chaudieres REP

Energy Technology Data Exchange (ETDEWEB)

Sabaton, M.; Morilhat, P.; Savoldelli, D.; Genette, P.

1995-10-01

Electricite de France (EDF), the french national electricity company, is operating 54 standardized pressurizer water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYsteme de Surveillance en FAtigue de la Chaudiere) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After the first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF`s nuclear 900 MW is planned after one more year of feedback experience. (authors). 12 refs., 3 figs.

Experimental and numerical investigation of the coolant mixing during fast deboration transients

International Nuclear Information System (INIS)

Hoehne, T.; Rohde, U.; Weiss, F.P.

1999-01-01

For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the VVER-440 were performed with a CFD code (CFX-4). The comparison with experimental data and an analytical mixing model which is implemented in the neutron-kinetic code DYN3D showed a good agreement for near-nominal conditions. First experiments at the Rossendorf Mixing Test Facility ROCOM were performed simulating the start-up of the first main coolant pump. The reference reactor for the geometrically 1:5 scaled Plexiglas model is the German Konvoi type PWR. After demonstrating the capability of the CFD code to simulate these complicated flow transients, calculations were performed for the start-up of the first pump in a VVER-440 type reactor. These calculations are a first step of understanding the coolant mixing in the RPV of a VVER-440 type reactor under transient conditions. The results of the calculation show a very complex flow in the downcomer. A high downcomer of VVER-440 and the existence of the lower control rod chamber support coolant mixing is concluded. (author)

Numerical simulation of CALIBAN reactivity perturbation experiments using neptunium-237 samples

International Nuclear Information System (INIS)

Humbert, Philippe; Mechitoua, Boukhmes

2003-01-01

In order to contribute to the validation of nuclear data used in critical mass computation, reactivity perturbation experiments using 237 Np samples have been performed at CEA-Valduc using the fast pulsed reactor CALIBAN operated in continuous mode. In this paper we report these experiments together with the numerical calculations. The calculations were carried out using PANDA, a S N code developed at CEA-Bruyeres-le-Chatel for classic criticality and stochastic neutronics applications and with MCNP4C which is a commonly used Monte Carlo code. A good agreement was found between experimental and deterministic results. (author)

New developments in French transient monitoring: SYSFAC

International Nuclear Information System (INIS)

L'huby, Y.; Genette, P.; Faidy, C.; Kappler, F.; Balley, J.; Bimont, G.

1991-01-01

After more than ten years of experience with Transient Monitoring and Logging Procedure (TMLP) and six years of successfully experience with Fatiguemeters, EDF has decided to study a new concept of Fatigue Monitoring System: SYSFAC. This new automatic system which is developed to be operating in all the French PWR units is composed of three modules: mechanical transient logging, functional transient logging and fatiguemeters. This application must be connected to the on-site data acquisition system without complementary instrumentation on the plant. (author)

Rapid and transient stimulation of intracellular reactive oxygen species by melatonin in normal and tumor leukocytes

International Nuclear Information System (INIS)

Radogna, Flavia; Paternoster, Laura; De Nicola, Milena; Cerella, Claudia; Ammendola, Sergio; Bedini, Annalida; Tarzia, Giorgio; Aquilano, Katia; Ciriolo, Maria; Ghibelli, Lina

2009-01-01

Melatonin is a modified tryptophan with potent biological activity, exerted by stimulation of specific plasma membrane (MT1/MT2) receptors, by lower affinity intracellular enzymatic targets (quinone reductase, calmodulin), or through its strong anti-oxidant ability. Scattered studies also report a perplexing pro-oxidant activity, showing that melatonin is able to stimulate production of intracellular reactive oxygen species (ROS). Here we show that on U937 human monocytes melatonin promotes intracellular ROS in a fast (< 1 min) and transient (up to 5-6 h) way. Melatonin equally elicits its pro-radical effect on a set of normal or tumor leukocytes; intriguingly, ROS production does not lead to oxidative stress, as shown by absence of protein carbonylation, maintenance of free thiols, preservation of viability and regular proliferation rate. ROS production is independent from MT1/MT2 receptor interaction, since a) requires micromolar (as opposed to nanomolar) doses of melatonin; b) is not contrasted by the specific MT1/MT2 antagonist luzindole; c) is not mimicked by a set of MT1/MT2 high affinity melatonin analogues. Instead, chlorpromazine, the calmodulin inhibitor shown to prevent melatonin-calmodulin interaction, also prevents melatonin pro-radical effect, suggesting that the low affinity binding to calmodulin (in the micromolar range) may promote ROS production.

A fast reactor transient analysis methodology for personal computers

International Nuclear Information System (INIS)

Ott, K.O.

1993-01-01

A simplified model for a liquid-metal-cooled reactor (LMR) transient analysis, in which point kinetics as well as lumped descriptions of the heat transfer equations in all components are applied, is converted from a differential into an integral formulation. All 30 differential balance equations are implicitly solved in terms of convolution integrals. The prompt jump approximation is applied as the strong negative feedback effectively keeps the net reactivity well below prompt critical. After implicit finite differencing of the convolution integrals, the kinetics equation assumes a new form, i.e., the quadratic dynamics equation. In this integral formulation, the initial value problem of typical LMR transients can be solved with large item steps (initially 1 s, later up to 256 s). This then makes transient problems amenable to a treatment on personal computer. The resulting mathematical model forms the basis for the GW-BASIC program LMR transient calculation (LTC) program. The LTC program has also been converted to QuickBASIC. The running time for a 10-h transient overpower transient is then ∼40 to 10 s, depending on the hardware version (286, 386, or 486 with math coprocessors)

Results of LWR core transient benchmarks

International Nuclear Information System (INIS)

Finnemann, H.; Bauer, H.; Galati, A.; Martinelli, R.

1993-10-01

LWR core transient (LWRCT) benchmarks, based on well defined problems with a complete set of input data, are used to assess the discrepancies between three-dimensional space-time kinetics codes in transient calculations. The PWR problem chosen is the ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal - hydraulic models of the codes. The 63 sets of submitted solutions are analyzed by comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations. The BWR problems considered are reactivity excursions caused by cold water injection and pressurization events. In the present paper, only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution the evaluation of the 8 sets of BWR contributions relies on a synthetic comparative discussion. The results of this first phase of LWRCT benchmark calculations are quite satisfactory, though there remain some unresolved issues. It is therefore concluded that even more challenging problems can be successfully tackled in a suggested second test phase. (authors). 46 figs., 21 tabs., 3 refs

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

Measurements of fuel temperature coefficient of reactivity on a commercial AGR

International Nuclear Information System (INIS)

Telford, A.; Bridge, M.J.

1978-01-01

Tests have been carried out on the commercial AGR at Hikley Point to determine the fuel temperature coefficient of reactivity, an important safety related parameter. Reactor neutron flux was measured during transients induced by movement of a bank of control rods from one steady position to another. An inverse kinetics analysis was applied to the measured flux to determine the change which occured in core reactivity as the fuel temperature changed. The variation of mean fuel temperature was deduced from the flux transient by means of a nine-plane thermal hydraulics representation of the AGR fuel channel. Results so far obtained confirm the predicted variation of fuel temperature coefficient with butn-up. (author)

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

1980-05-01

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin.

Burst-suppression is reactive to photic stimulation in comatose children with acquired brain injury

DEFF Research Database (Denmark)

Nita, Dragos A.; Moldovan, Mihai; Sharma, Roy

2016-01-01

reactivity. We quantified reactivity by measuring the change in the burst ratio (fraction of time in burst) following photic stimulation. Results: Photic stimulation evoked bursts in all patients, resulting in a transient increase in the burst ratio, while the mean heart rate remained unchanged......Objective: Burst-suppression is an electroencephalographic pattern observed during coma. In individuals without known brain pathologies undergoing deep general anesthesia, somatosensory stimulation transiently increases the occurrence of bursts. We investigated the reactivity of burst......-suppression in children with acquired brain injury. Methods: Intensive care unit electroencephalographic monitoring recordings containing burst-suppression were obtained from 5 comatose children with acquired brain injury of various etiologies. Intermittent photic stimulation was performed at 1 Hz for 1 min to assess...

Predictive modeling of transient storage and nutrient uptake: Implications for stream restoration

Science.gov (United States)

O'Connor, Ben L.; Hondzo, Miki; Harvey, Judson W.

2010-01-01

This study examined two key aspects of reactive transport modeling for stream restoration purposes: the accuracy of the nutrient spiraling and transient storage models for quantifying reach-scale nutrient uptake, and the ability to quantify transport parameters using measurements and scaling techniques in order to improve upon traditional conservative tracer fitting methods. Nitrate (NO3–) uptake rates inferred using the nutrient spiraling model underestimated the total NO3– mass loss by 82%, which was attributed to the exclusion of dispersion and transient storage. The transient storage model was more accurate with respect to the NO3– mass loss (±20%) and also demonstrated that uptake in the main channel was more significant than in storage zones. Conservative tracer fitting was unable to produce transport parameter estimates for a riffle-pool transition of the study reach, while forward modeling of solute transport using measured/scaled transport parameters matched conservative tracer breakthrough curves for all reaches. Additionally, solute exchange between the main channel and embayment surface storage zones was quantified using first-order theory. These results demonstrate that it is vital to account for transient storage in quantifying nutrient uptake, and the continued development of measurement/scaling techniques is needed for reactive transport modeling of streams with complex hydraulic and geomorphic conditions.

Experiments on transient melting of tungsten by ELMs in ASDEX Upgrade

Science.gov (United States)

Krieger, K.; Balden, M.; Coenen, J. W.; Laggner, F.; Matthews, G. F.; Nille, D.; Rohde, V.; Sieglin, B.; Giannone, L.; Göths, B.; Herrmann, A.; de Marne, P.; Pitts, R. A.; Potzel, S.; Vondracek, P.; ASDEX-Upgrade Team; EUROfusion MST1 Team

2018-02-01

Repetitive melting of tungsten by power transients originating from edge localized modes (ELMs) has been studied in ASDEX Upgrade. Tungsten samples were exposed to H-mode discharges at the outer divertor target plate using the divertor manipulator II (DIM-II) system (Herrmann et al 2015 Fusion Eng. Des. 98-9 1496-9). Designed as near replicas of the geometries used also in separate experiments on the JET tokamak (Coenen et al 2015 J. Nucl. Mater. 463 78-84 Coenen et al 2015 Nucl. Fusion 55 023010; Matthews et al 2016 Phys. Scr. T167 7), the samples featured a misaligned leading edge and a sloped ridge respectively. Both structures protrude above the default target plate surface thus receiving an increased fraction of the parallel power flux. Transient melting by ELMs was induced by moving the outer strike point to the sample location. The temporal evolution of the measured current flow from the samples to vessel potential confirmed transient melting. Current magnitude and dependency from surface temperature provided strong evidence for thermionic electron emission as main origin of the replacement current driving the melt motion. The different melt patterns observed after exposures at the two sample geometries support the thermionic electron emission model used in the MEMOS melt motion code, which assumes a strong decrease of the thermionic net current at shallow magnetic field to surface angles (Pitts et al 2017 Nucl. Mater. Energy 12 60-74). Post exposure ex situ analysis of the retrieved samples show recrystallization of tungsten at the exposed surface areas to a depth of up to several mm. The melt layer transport to less exposed surface areas leads to ratcheting pile up of re-solidified debris with zonal growth extending from the already enlarged grains at the surface.

Preliminary analysis of typical transients in fusion driven subcritical system (FDS-I)

International Nuclear Information System (INIS)

Bai Yunqing; Ke Yan; Wu Yican

2007-01-01

The potential safety characteristic is expected as one of the advantages of fusion-driven subcritical system (FDS-I) for the transmutation and incineration of nuclear waste compared with the critical reactor. Transients of the FDS-I may occur due to the perturbation of external neutron source, the failure of functional device, and the occurrence of the uncontrolled event. As typical transient scenarios, the following cases were analyzed: unprotected plasma overpower (UPOP), unprotected loss of flow (ULOF), unprotected transient overpower (UTOP). The transient analyses for the FDS-I were performed with a coupled two-dimensional thermal-hydraulics and neutronics transient analysis code NTC2D. The negative feedback of reactivity is the interesting safety feature of FDS-I as temperature increase, due to the fuel form of the circulating particle. The present simulation results showed that the current FDS-I design has a resistance against severe transient scenarios. (author)

A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

Energy Technology Data Exchange (ETDEWEB)

Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

Energy Technology Data Exchange (ETDEWEB)

Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

Transient wave behaviour over an underwater sliding hump from experiments and analytical and numerical modelling

Energy Technology Data Exchange (ETDEWEB)

Callaghan, David P.; Nielsen, Peter [The University of Queensland, School of Civil Engineering, Brisbane (Australia); Ahmadi, Afshin [Kellogg Brown and Root Pty Ltd, Brisbane, QLD (Australia)

2011-12-15

Flume measurements of a one-dimensional sliding hump starting from rest in quiescence fresh water indicate that when the hump travels at speed less than the shallow-water wave celerity, three waves emerge, travelling in two directions. One wave travels in the opposite direction to the sliding hump at approximately the shallow-water wave celerity (backward free wave). Another wave travels approximately in step with the hump (forced wave), and the remaining wave travels in the direction of the hump at approximately the shallow-water wave celerity (forward free wave). These experiments were completed for a range of sliding hump speed relative to the shallow-water wave celerity, up to unity of this ratio, to investigate possible derivation from solutions of the Euler equation with non-linear and non-hydrostatic terms being included or excluded. For the experimental arrangements tested, the forced waves were negative (depression or reduced water surface elevation) waves while the free waves were positive (bulges or increased water surface elevation). For experiments where the sliding hump travelled at less than 80% of the shallow-water wave celerity did not include transient behaviour measurements (i.e. when the three waves still overlapped). The three wave framework was partially supported by these measurements in that the separated forward and forced waves were compared to measurements. For the laboratory scale experiments, the forward free wave height was predicted reasonably by the long-wave equation (ignoring non-linear and non-hydrostatic terms) when the sliding hump speed was less than 80% of the shallow-water wave celerity. The forced wave depression magnitude required the Euler equations for all hump speed tested. The long-wave solution, while being valid in a limited parameter range, does predict the existence of the three waves as found in these experiments (forward travelling waves measured quantitatively while the backward travelling waves visually by video

BARS - a heterogeneous code for 3D pin-by-pin LWR steady-state and transient calculation

International Nuclear Information System (INIS)

Avvakumov, A.V.; Malofeev, V.M.

2000-01-01

A 3D pin-by-pin dynamic model for LWR detailed calculation was developed. The model is based on a coupling of the BARS neutronic code with the RELAP5/MOD3.2 thermal hydraulic code. This model is intended to calculate a fuel cycle, a xenon transient, and a wide range of reactivity initiated accidents in a WWER and a PWR. Galanin-Feinberg heterogeneous method was realized in the BARS code. Some results for a validation of the heterogeneous method are presented for reactivity coefficients, a pin-by-pin power distribution, and a fast pulse transient. (Authors)

Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

International Nuclear Information System (INIS)

Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

1997-01-01

Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel

Clopidogrel discontinuation and platelet reactivity following coronary stenting

LENUS (Irish Health Repository)

2011-01-01

Summary. Aims: Antiplatelet therapy with aspirin and clopidogrel is recommended for 1 year after drug-eluting stent (DES) implantation or myocardial infarction. However, the discontinuation of antiplatelet therapy has become an important issue as recent studies have suggested a clustering of ischemic events within 90 days of clopidogrel withdrawal. The objective of this investigation was to explore the hypothesis that there is a transient ‘rebound’ increase in platelet reactivity within 3 months of clopidogrel discontinuation. Methods and Results: In this prospective study, platelet function was assessed in patients taking aspirin and clopidogrel for at least 1 year following DES implantation. Platelet aggregation was measured using a modification of light transmission aggregometry in response to multiple concentrations of adenosine diphosphate (ADP), epinephrine, arachidonic acid, thrombin receptor activating peptide and collagen. Clopidogrel was stopped and platelet function was reassessed 1 week, 1 month and 3 months later. Thirty-two patients on dual antiplatelet therapy were recruited. Discontinuation of clopidogrel increased platelet aggregation to all agonists, except arachidonic acid. Platelet aggregation in response to ADP (2.5, 5, 10, and 20 μm) and epinephrine (5 and 20 μm) was significantly increased at 1 month compared with 3 months following clopidogrel withdrawal. Thus, a transient period of increased platelet reactivity to both ADP and epinephrine was observed 1 month after clopidogrel discontinuation. Conclusions: This study demonstrates a transient increase in platelet reactivity 1 month after clopidogrel withdrawal. This phenomenon may, in part, explain the known clustering of thrombotic events observed after clopidogrel discontinuation. This observation requires confirmation in larger populations.

Experimental evaluation of reactivity constraints for the closed-loop control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.; Lanning, D.D.; Ray, A.

1984-01-01

General principles for the closed-loop, digital control of reactor power have been identified, quantitatively enumerated, and experimentally demonstrated on the 5 MWt Research Reactor, MITR-II. The basic concept is to restrict the net reactivity so that it is always possible to make the reactor period infinite at the desired termination point of a transient by reversing the direction of motion of whatever control mechanism is associated with the controller. This capability is formally referred to as ''feasibility of control''. A series of ten experiments have been conducted over a period of eighteen months to demonstrate the efficacy of this property for the automatic control of reactor power. It has been shown that a controller which possesses this property is capable of both raising and lowering power in a safe, efficient manner while using a control rod of varying differential worth, that the reactivity constraints are a sufficient condition for the automatic control of reactor power, and that the use of a controller based on reactivity constraints can prevent overshoots either due to attempts to control a transient with a control rod of insufficient differential worth or due to failure to properly estimate when to commence rod insertion. Details of several of the more significant tests are presented together with a discussion of the rationale for the development of closed-loop control in large commercial power systems. Specific consideration is given to the motivation for designing a controller based on feasibility of control and the associated licensing issues

Entropy-based critical reaction time for mixing-controlled reactive transport

DEFF Research Database (Denmark)

Chiogna, Gabriele; Rolle, Massimo

2017-01-01

Entropy-based metrics, such as the dilution index, have been proposed to quantify dilution and reactive mixing in solute transport problems. In this work, we derive the transient advection dispersion equation for the entropy density of a reactive plume. We restrict our analysis to the case where...... the concentration distribution of the transported species is Gaussian and we observe that, even in case of an instantaneous complete bimolecular reaction, dilution caused by dispersive processes dominates the entropy balance at early times and results in the net increase of the entropy density of a reactive species...

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)

Instrumented anvil-on-rod impact experiments for validating constitutive strength model for simulating transient dynamic deformation response of metals

International Nuclear Information System (INIS)

Martin, M.; Shen, T.; Thadhani, N.N.

2008-01-01

Instrumented anvil-on-rod impact experiments were performed to access the applicability of this approach for validating a constitutive strength model for dynamic, transient-state deformation and elastic-plastic wave interactions in vanadium, 21-6-9 stainless steel, titanium, and Ti-6Al-4V. In addition to soft-catching the impacted rod-shaped samples, their transient deformation states were captured by high-speed imaging, and velocity interferometry was used to record the sample back (free) surface velocity and monitor elastic-plastic wave interactions. Simulations utilizing AUTODYN-2D hydrocode with Steinberg-Guinan constitutive equation were used to generate simulated free surface velocity traces and final/transient deformation profiles for comparisons with experiments. The simulations were observed to under-predict the radial strain for bcc vanadium and fcc steel, but over-predict the radial strain for hcp titanium and Ti-6Al-4V. The correlations illustrate the applicability of the instrumented anvil-on-rod impact test as a method for providing robust model validation based on the entire deformation event, and not just the final deformed state

Inherent safety that the reactivity effect of core bending in fast reactors brings about

International Nuclear Information System (INIS)

Nakagawa, Masatoshi; Yagawa, Genki.

1994-01-01

FBRs have the merit on safety by low operation pressure and the large heat capacity of coolant, in addition, due to the core temperature rise at the time of accidents and the thermal expansion of core structures, the negative feedback of reactivity can be expected. Recently, attention has been paid to the negative feedback of reactivity due to core bending. It can be expected also in the core of limited free bow type. Bending is caused by the difference of thermal expansion on six surfaces of hexagonal wrapper tubes. The bending changes core reactivity and exerts effects to fuel exchange force and operation, insertion of control rods and the structural soundness of fuel assemblies. for the purpose of limiting the effect that core bending exerts to core characteristics to allowable range, core constraint mechanism is installed. The behavior of core bending at the time of anticipated transient without scram is explained. The example of the analysis of PRISM reactor is shown. The experiment that confirmed the negative feedback of reactivity due to core bending under the condition of ULOF was that at the fast flux test facility. (K.I.)

Advanced Instrumentation for Transient Reactor Testing

Energy Technology Data Exchange (ETDEWEB)

Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

2018-01-31

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

Various reactivity effects value for assuring fast reactor core inherent safety

International Nuclear Information System (INIS)

Belov, S.B.; Vasilyev, B.A.

1991-01-01

The paper presents the results of temperature and power reactivity feedback components calculations for fast reactors with different core volume when using oxide, carbide, nitride and metal fuel. Reactor parameters change in loss of flow without scram and transient over power without scram accidents was evaluated. The importance of various reactivity feedback components in restricting the consequences of these accidents has been analyzed. (author)

Learning from anticipated and abnormal plant transients

International Nuclear Information System (INIS)

Varnado, B.

1983-01-01

A report is given of the American Nuclear Society topical meeting on Anticipated and Abnormal Transients in Light Water Reactors held in Jackson, Wyoming in September 1983. Industry involvement in the evaluation of operating experience, human error contributions, transient management, thermal hydraulic modelling, the role of probabilistic risk assessment and the cost of transient incidents are discussed. (U.K.)

Improved Reactive Flow Modeling of the LX-17 Double Shock Experiments

Science.gov (United States)

Rehagen, Thomas J.; Vitello, Peter

2017-06-01

Over driven double shock experiments provide a measurement of the properties of the reaction product states of the insensitive high explosive LX-17 (92.5% TATB and 7.5% Kel-F by weight). These experiments used two flyer materials mounted on the end of a projectile to send an initial shock through the LX-17, followed by a second shock of a higher magnitude into the detonation products. In the experiments, the explosive was initially driven by the flyer plate to pressures above the Chapman-Jouguet state. The particle velocity history was recorded by Photonic Doppler Velocimetry (PDV) probes pointing at an aluminum foil coated LiF window. The PDV data shows a sharp initial shock and decay, followed by a rounded second shock. Here, the experimental results are compared to 2D and 3D Cheetah reactive flow modeling. Our default Cheetah reactive flow model fails to accurately reproduce the decay of the first shock or the curvature or strength of the second shock. A new model is proposed in which the carbon condensate produced in the reaction zone is controlled by a kinetic rate. This allows the carbon condensate to be initially out of chemical equilibrium with the product gas. This new model reproduces the initial detonation peak and decay, and matches the curvature of the second shock, however, it still over-predicts the strength of the second shock. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract No. DE-AC52-07NA27344.

Future Transient Testing of Advanced Fuels

Energy Technology Data Exchange (ETDEWEB)

Jon Carmack

2009-09-01

the refurbishment and restart of TREAT. •TREAT is an absolute necessity in the suite of reactor fuel test capabilities •TREAT yields valuable information on reactivity effects, margins to failure, fuel dispersal, and failure propagation •Most importantly, interpretation of TREAT experiment results is a stringent test of the integrated understanding of fuel performance.

Future Transient Testing of Advanced Fuels

International Nuclear Information System (INIS)

Carmack, Jon

2009-01-01

refurbishment and restart of TREAT: (1) TREAT is an absolute necessity in the suite of reactor fuel test capabilities; (2) TREAT yields valuable information on reactivity effects, margins to failure, fuel dispersal, and failure propagation; and (3) Most importantly, interpretation of TREAT experiment results is a stringent test of the integrated understanding of fuel performance.

Parametric study of postulated reactivity transients due to ingress of heavy water from the reflector tank into the converted core of APSARA reactor

International Nuclear Information System (INIS)

Sankaranarayanan, S.

2004-01-01

Research reactors in the power range 5-10 MW with useable neutron flux values >1.OE+14 n/sqcm/sec can be constructed using LEU fuel with light water for neutron moderation and fuel cooling. In order to obtain a large irradiation volume, a heavy water reflector is used where fairly high neutron flux levels can be obtained. A prototype LEU fuelled 5/10 MW reactor design has been developed in the Bhabha Atomic Research Centre in Trombay. Work is on hand to carry out technology simulation of this reactor design by converting the pool type reactor APSARA in BARC. Presently the Apsara reactor uses MTh type high enriched U-Al alloy plate type fuel loaded in a 7x7 grid with a square lattice pitch of 76.8 mm. The reactor has three control-scram-shut off rods and one regulating control rod. In the first phase of the simulation studies, it is proposed to use the existing high enriched uranium fuel in a modified core with 37 positions arranged with a square lattice pitch of 84.8 mm, surrounded by a 50 cm thick heavy water reflector. Subsequently the converted core will use plate-type low enriched uranium suicide fuel. One of the accident scenarios postulated for the safety evaluation of the modified APSARA reactor is the reactivity transient due to the ingress of heavy water into the core through a small sized rupture in the aluminium wall of the reflector tank. Parametric analyses were done for the safety evaluation of modified Apsara reactor, for postulated leak of heavy water into the core from the reflector tank. A simplified computer code REDYN, based on point model reactor kinetics with one effective group of delayed neutrons is used for the analyses. Results of several parametric cases used in the study show that it is possible to contain the consequences of this type of reactivity transient within acceptable fuel and coolant thermal safety limits

Predictive Modeling of Transient Storage and Nutrient Uptake: Implications for Stream Restoration

Energy Technology Data Exchange (ETDEWEB)

O’Connor, Ben L.; Hondzo, Miki; Harvey, Judson W.

2010-12-01

This study examined two key aspects of reactive transport modeling for stream restoration purposes: the accuracy of the nutrient spiraling and transient storage models for quantifying reach-scale nutrient uptake, and the ability to quantify transport parameters using measurements and scaling techniques in order to improve upon traditional conservative tracer fitting methods. Nitrate (NO-3)(NO3-) uptake rates inferred using the nutrient spiraling model underestimated the total NO-3NO3- mass loss by 82%, which was attributed to the exclusion of dispersion and transient storage. The transient storage model was more accurate with respect to the NO-3NO3- mass loss (±20%) and also demonstrated that uptake in the main channel was more significant than in storage zones. Conservative tracer fitting was unable to produce transport parameter estimates for a riffle-pool transition of the study reach, while forward modeling of solute transport using measured/scaled transport parameters matched conservative tracer breakthrough curves for all reaches. Additionally, solute exchange between the main channel and embayment surface storage zones was quantified using first-order theory. These results demonstrate that it is vital to account for transient storage in quantifying nutrient uptake, and the continued development of measurement/scaling techniques is needed for reactive transport modeling of streams with complex hydraulic and geomorphic conditions.

Transient plasma cobalamin elevation in patients with pneumonia - two case reports

DEFF Research Database (Denmark)

Rahbek, Martin Torp; Scheller, Rudolf; Nybo, Mads

2018-01-01

We report two cases of transient significantly elevated plasma cobalamin (B12) in geriatric patients acutely admitted with fever, increased C-reactive protein and X-ray verified pneumonia. Extensive diagnostic workup did not reveal kidney or liver disease, neither any signs of cancer. Furthermore...

Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

International Nuclear Information System (INIS)

Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

2013-01-01

The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

Fuel pin response to an overpower transient in an LMFBR

International Nuclear Information System (INIS)

Grosberg, A.J.; Head, J.L.

1979-01-01

This paper describes a method by which the ability of a whole-core code accurately to predict the time and location of the first fuel pin failures may be tested. The method involves the use of a relatively simple whole-core code to 'drive' a sophisticated fuel pin code, which is far too complex to be used within a whole-core code but which is potentially capable of modelling reliably the response of an individual fuel pin. The method cannot follow accurately the subsequent course of the transient because the simple whole-core code does not model the reactivity effects of events which may follow pin failure. The codes used were the simple whole-core code FUTURE and the fuel pin behaviour code FRUMP. The paper describes an application of the method to analyse a hypothetical LMFBR accident in which the control rods were assumed to be driven from the core at maximum speed, with all trip circuits failed. Taking 0.5% clad strain as a clad failure criterion, failure was predicted to occur at the top of the active core at about 10s into the transient. A repeat analysis, using an alternative clad yield criterion which is thought to be more realistic, indicated failure at the same position but 24s into the transient. This is after the onset of sodium boiling. Pin failure at the top of the core are likely to cause negative reactivity changes. In this hypothetical accident, pin failures are likely, therefore, to have a moderating effect on the course of the transient. (orig.)

CO2 injection into fractured peridotites: a reactive percolation experiment

Science.gov (United States)

Escario, S.; Godard, M.; Gouze, P.; Leprovost, R.; Luquot, L.; Garcia-Rios, M.

2017-12-01

Mantle peridotites have the potential to trap CO2 as carbonates. This process observed in ophiolites and in oceanic environments provides a long term and safe storage for CO2. It occurs as a part of a complex suite of fluid-rock reactions involving silicate dissolution and precipitation of hydrous phases, carbonates and minor phases that may in turn modify the hydrodynamic properties and the reactivity of the reacted rocks. The efficiency and lastingness of the process require the renewal of fluids at the mineral-fluid interface. Fractures are dominant flow paths in exhumed mantle sections. This study aims at better understanding the effect of CO2-enriched saline fluids on hydrodynamic and chemical processes through fractured peridotites. Experiments were performed using the reactive percolation bench ICARE Lab 3 - Géosciences Montpellier. It allows monitoring the permeability changes during experiments. Effluents are recurrently sampled for analysing cation concentration, pH and alkalinity. Reacted rock samples were characterized by high resolution X-ray microtomography (ESRF ID19, Grenoble, France) and SEM. Experiments consisted in injecting CO2-enriched brines (NaCl 0.5 M) at a rate of 6 mL.h-1 into artificially fractured cores (9 mm diameter × 20 mm length) of Oman harzburgites at T=170°C and Ptotal = 25 MPa for up to 2 weeks. Fractures are of few µm apertures with rough walls. Three sets of experiments were performed at increasing value of [CO2] (0, 0.1 and 1 mol/kg). All experiments showed a decrease in permeability followed by steady state regime that can be caused by a decrease in the roughness of fracture walls (dissolution dominated process), thus favouring fracture closing, or by the precipitation of secondary phases. Maximum enrichments in Mg, Fe and Ca of the effluent fluids occur during the first 2 hours of the experiments whereas Si displays a maximum enrichment at t = 20 h, suggesting extensive dissolution. Maximum enrichments are observed with

Evaluating advanced LMR [liquid metal reactor] reactivity feedbacks using SSC

International Nuclear Information System (INIS)

Slovik, G.C.; Van Tuyle, G.J.; Kennett, R.J.; Cheng, H.S.

1988-01-01

Analyses of the PRISM and SAFR Liquid Metal Reactors with SSC are discussed from a safety and licensing perspective. The PRISM and SAFR reactors with metal fuel are designed for inherent shutdown responses to loss-of-flow and loss-of-heat-sink events. The demonstration of this technology was performed by EBR-II during experiments in April 1986 by ANL (Planchon, et al.). Response to postulated TOPs (control rod withdrawal) are made acceptable largely by reducing reactivity swings, and therefore minimizing the size of possible ractivity insertions. Analyses by DOE and the contractors GE, RI, and ANL take credit for several reactivity feedback mechanisms during transient calculations. These feedbacks include Doppler, sodium density, and thermal expansion of the grid plates, the load pads, the fuel (axial) and the control rod which are now factored into the BNL SSC analyses. The bowing feedback mechanism is not presently modeled in the SSC due to its complexity and subsequent large uncertainty. The analysis is conservative by not taking credit for this negative feedback mechanism. Comparisons of BNL predictions with DOE contractors are provided

Mechanisms of ignition by transient energy deposition: Regimes of combustion wave propagation

OpenAIRE

Kiverin, A. D.; Kassoy, D. R.; Ivanov, M. F.; Liberman, M. A.

2013-01-01

Regimes of chemical reaction wave propagating in reactive gaseous mixtures, whose chemistry is governed by chain-branching kinetics, are studied depending on the characteristics of a transient thermal energy deposition localized in a finite volume of reactive gas. Different regimes of the reaction wave propagation are initiated depending on the amount of deposited thermal energy, power of the source, and the size of the hot spot. The main parameters which define regimes of the combustion wave...

The reactor kinetics code tank: a validation against selected SPERT-1b experiments

International Nuclear Information System (INIS)

Ellis, R.J.

1990-01-01

The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

Reactivity to Social Stress in Subclinical Social Anxiety: Emotional Experience, Cognitive Appraisals, Behavior, and Physiology

Science.gov (United States)

Crişan, Liviu G.; Vulturar, Romana; Miclea, Mircea; Miu, Andrei C.

2016-01-01

Recent research indicates that subclinical social anxiety is associated with dysfunctions at multiple psychological and biological levels, in a manner that seems reminiscent of social anxiety disorder (SAD). This study aimed to describe multidimensional responses to laboratory-induced social stress in an analog sample selected for social anxiety symptoms. State anxiety, cognitive biases related to negative social evaluation, speech anxiety behaviors, and cortisol reactivity were assessed in the Trier Social Stress Test (TSST). Results showed that social anxiety symptoms were associated with increased state anxiety, biased appraisals related to the probability and cost of negative social evaluations, behavioral changes in facial expression that were consistent with speech anxiety, and lower cortisol reactivity. In addition, multiple interrelations between responses in the TSST were found, with positive associations between subjective experience, cognitive appraisals, and observable behavior, as well as negative associations between each of the former two types of response and cortisol reactivity. These results show that in response to social stressors, subclinical social anxiety is associated with significant changes in emotional experience, cognitive appraisals, behaviors, and physiology that could parallel those previously found in SAD samples. PMID:26858658

Preliminary Assessment of Transient of Over Power Accident for DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

TRACE code was selected as one of candidates for audit code, so sodium properties and heat transfer model in the code was verified first. On the basis of MARS-LMR code input, DSFR-600 TRACE model was developed and applied to PHTS tube rupture case, one of design base events (DBE) of DSFR-600. In this study, Transients of Over Power (TOP) event is assessed using TRACE code as one another case of DBEs of DSFR-600 for preparation of audit calculation of PGSFR.One of the design base events, transients of over power of Demonstration Sodium cooled Fast Reactor was simulated using TRACE code. Predicted fuel temperature showed that the peak fuel temperature occurs when the reactor scrammed and predicted temperature was similar to the MARS-LMRs assessment by KAERI. In this study, it is found that the second peak of fuel temperature is influenced by the inventory of steam generator and the natural circulation characteristic of the reactor vessel pool. Pre-calculation of the unprotected transients of over power with conservative reactivity assumption showed that this assumption is conservative in design base even assessment. However the method of measurement and applying the core radial, fuel and control rod axial expansion reactivity feedback is crucial in BDBE assessment of SFR.

Experiment using TRACY and its research results

International Nuclear Information System (INIS)

Nakajima, Ken; Ono, Akio; Okazaki, Shuji

1997-01-01

Japan Atomic Energy Research Institute started a critical accident trial experiment since 1995 using TRACY (Transient critical experimental apparatus) installed in NUCEF, aiming to elucidate critical accident phenomenon in solution state nuclear fuel and to establish a rational critical accident evaluation method. The TRACY is an apparatus to conduct the experiment beyond critical (super critical) state using uranyl nitrate low condensed aqueous solution treated at reprocessing facility for its fuel. In the TRACY, aiming to evaluate 1) nuclear fission numbers at the burst output portion, total nuclear fission numbers, and maximum nuclear fission ratio (peak output) and pressure, the following conditions and data are required for analysis and evaluation of them at a supposed critical accident: a) system conditions, b) initial conditions, c) nuclear and thermal constants, d) reactivity addition conditions, e) reactivity feed-back mechanism, and f) mobilities of main isotopes. In this paper, experimental plan, summary of experimental apparatus, the obtained results, and future planning of the TRACY were described. (G.K.)

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.

Internal fuel motion as an inherent shutdown mechanism for LMFBR accidents: PINEX-3, PINEX-2, and HUT 5-2A experiments

International Nuclear Information System (INIS)

Ferrell, P.C.; Porten, D.R.; Martin, F.J.

1981-01-01

The PINEX-2 experiment verified the concept of axial internal molten fuel motion within annular fuel, representing an inherent shutdown mechanism for hypothetical transient overpower excursions on the order of 5$/s. The PINEX-3 experiment, simulating a 50 cents/s transient overpower, showed that limitations on the effectiveness of fuel motion may arise from freezing of the fuel and blockage of the internal movement. Analysis of these experiments was performed to assess the physical processes that dominate fuel relocation potential and to apply them to prototypic LMFBR pin conditions. Results indicate that internal fuel motion should be reliable as a shutdown mechanism in LMFBR's for a range of reactivity insertion rates beyond presently available experimental data

Visual investigation of transient fuel behavior under a rapid heating condition

International Nuclear Information System (INIS)

Saito, Shinzo

1981-10-01

An in-reactor experimental research on fuel behavior under reactivity initiated accident (RIA) conditions is being conducted in the Nuclear Safety Research Reactor (NSRR). The optical system in which a non-browning lens periscope is directly installed in the test section was successfully developed for photographing transient fuel behavior. Several phenomena which had never been revealed before were observed in the slow motion pictures taken in the NSRR experiments which were performed in the water and air environments. As for incipient failure mechanism for an unirradiated fuel rod under RIA conditions, brittle fracture of the cladding during quenching is dominant. However, a split cracking possibly occurs during even red hot state of the cladding. It is considered that the crack is generated by the local internal pressure increase at the specified region blocked up due to the melting of the cladding inner surface. The film boiling is unexpectablly violent specially in the early stage of the transient, and film thickness becomes 5 -- 6 mm at maximum. The observed thick vapor film can not be explained by the conventional theory, but the effect of hydrogen which is produced by Zircaloy-water reaction is reasonably explained to form thick film in the report. The molten fuel was expelled from the cladding in the experiment which was performed in an air environment. The expelled fuel fragmented due to possibly initial motion effect, not mechanical collision effect, because Weber number is smaller than the critical value. (author)

Nuclear power plant transients: where are we

International Nuclear Information System (INIS)

Majumdar, D.

1984-05-01

This document is in part a postconference review and summary of the American Nuclear Society sponsored Anticipated and Abnormal Plant Transients in Light Water Reactors Conference held in Jackson, Wyoming, September 26-29, 1983, and in part a reflection upon the issues of plant transients and their impact on the viability of nuclear power. This document discusses state-of-the-art knowledge, deficiencies, and future directions in the plant transients area as seen through this conference. It describes briefly what was reported in this conference, emphasizes areas where it is felt there is confidence in the nuclear industry, and also discusses where the experts did not have a consensus. Areas covered in the document include major issues in operational transients, transient management, transient events experience base, the status of the analytical tools and their capabilities, probabilistic risk assessment applications in operational transients, and human factors impact on plant transients management

Investigation of the Buckling-Reactivity Conversion Coefficient using SRAC and MVP codes for UO2 Lattices in TCA experiments

International Nuclear Information System (INIS)

Le Dai Dien

2008-01-01

Benchmark experiments for International Reactor Physics Benchmark Experiments (IRPhE) Project carried out at TCA, the temperature effects on reactivity were studied for light water moderated and reflected UO 2 cores with/without soluble poisons. The buckling coefficient method using the measured critical water levels was proposed by Suzaki et al. The temperature dependence of buckling coefficient of reactivity and its variance by the core configurations of the benchmark experiments was investigated using SRAC and MVP calculations. From the calculations by SRAC as well as by MVP it is seen that the K-value can be taken as an average value only for each core with temperature changes which are considered as perturbation parameter. The difference between our calculations and benchmark results which uses constant K-value for all cores proves that the results depend on K-value and it play important role in defining reactivity effect using the water level worth method. (author)

SACI - O: A code for the analysis of transients in a pressurized water reactor core

International Nuclear Information System (INIS)

Resende Lobo, A.A. de; Soares, P.A.

1979-03-01

The SACI-O digital computer code consists basically of a pressurized water reactor core model. It is useful in the analysis of fast reactivity transients shorter than the loop transit time. The program can also be used for evaluating the core behaviour, during other transients, when the inlet coolant conditions are known. SACI-O uses point model neutron kinetics taking into account moderator and fuel reactivity effects, and fission products decay. The neutronic and thermal-hydraulic equations are solved for an average fuel pin described by a single axial node. To perform a more detailed calculation, the modeling of another cooling channel, which can be divided into axial segments, is included in the program. The reactor trip system is also partially simulated. (Author) [pt

Reactive Halogens in the Marine Boundary Layer (RHaMBLe): the tropical North Atlantic experiments

OpenAIRE

J. D. Lee; G. McFiggans; J. D. Allan; A. R. Baker; S. M. Ball; A. K. Benton; L. J. Carpenter; R. Commane; B. D. Finley; M. Evans; E. Fuentes; K. Furneaux; A. Goddard; N. Good; J. F. Hamilton

2010-01-01

The NERC UK SOLAS-funded Reactive Halogens in the Marine Boundary Layer (RHaMBLe) programme comprised three field experiments. This manuscript presents an overview of the measurements made within the two simultaneous remote experiments conducted in the tropical North Atlantic in May and June 2007. Measurements were made from two mobile and one ground-based platforms. The heavily instrumented cruise D319 on the RRS Discovery from Lisbon, Portugal to São Vicente, Cape Verde and back to Falmouth...

Transient or permanent fisheye views

DEFF Research Database (Denmark)

Jakobsen, Mikkel Rønne; Hornbæk, Kasper

2012-01-01

Transient use of information visualization may support specific tasks without permanently changing the user interface. Transient visualizations provide immediate and transient use of information visualization close to and in the context of the user’s focus of attention. Little is known, however......, about the benefits and limitations of transient visualizations. We describe an experiment that compares the usability of a fisheye view that participants could call up temporarily, a permanent fisheye view, and a linear view: all interfaces gave access to source code in the editor of a widespread...... programming environment. Fourteen participants performed varied tasks involving navigation and understanding of source code. Participants used the three interfaces for between four and six hours in all. Time and accuracy measures were inconclusive, but subjective data showed a preference for the permanent...

HECTR [Hydrogen Event: Containment Transient Response] analyses of the Nevada Test Site (NTS) premixed combustion experiments

International Nuclear Information System (INIS)

Wong, C.C.

1988-11-01

The HECTR (Hydrogen Event: Containment Transient Response) computer code has been developed at Sandia National Laboratories to predict the transient pressure and temperature responses within reactor containments for hypothetical accidents involving the transport and combustion of hydrogen. Although HECTR was designed primarily to investigate these phenomena in LWRs, it may also be used to analyze hydrogen transport and combustion experiments as well. It is in this manner that HECTR is assessed and empirical correlations, such as the combustion completeness and flame speed correlations for the hydrogen combustion model, if necessary, are upgraded. In this report, we present HECTR analyses of the large-scale premixed hydrogen combustion experiments at the Nevada Test Site (NTS) and comparison with the test results. The existing correlations in HECTR version 1.0, under certain conditions, have difficulty in predicting accurately the combustion completeness and burn time for the NTS experiments. By combining the combustion data obtained from the NTS experiments with other experimental data (FITS, VGES, ACUREX, and Whiteshell), a set of new and better combustion correlations was generated. HECTR prediction of the containment responses, using a single-compartment model and EPRI-provided combustion completeness and burn time, compares reasonably well against the test results. However, HECTR prediction of the containment responses using a multicompartment model does not compare well with the test results. This discrepancy shows the deficiency of the homogeneous burning model used in HECTR. To overcome this deficiency, a flame propagation model is highly recommended. 16 refs., 84 figs., 5 tabs

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)

Control of ZrH reactor reactivity perturbations during orbital maneuvers

International Nuclear Information System (INIS)

Audette, R.F.

1970-01-01

Scheduled and inadvertent vehicle maneuvers in manned and unmanned space missions may result in reactivity perturbations to the ZrH reactor due to fuel and control drum motion from acceleration forces. Potential power and outlet coolant temperature excursions could result in interruptions of PCS power generation, or excessive coolant temperatures if uncontrolled. This analysis compares potential uncontrolled reactor transients with allowable transients for uninterrupted electrical power generation from a Brayton system, and presents a control scheme to limit transient reactor outlet temperatures to 1250 0 F for a system designed to operate at a nominal 1200 0 F reactor outlet. Potential uncontrolled transients could result in a reactor outlet temperature swing of +-77 0 F about a nominal 1200 0 F and a reactor power swing of +92 Kwt and -67 Kwt about a nominal 130 Kwt for the Brayton System. (U.S.)

Particle beam experiments for the analysis of reactive sputtering processes in metals and polymer surfaces

Energy Technology Data Exchange (ETDEWEB)

Corbella, Carles; Grosse-Kreul, Simon; Kreiter, Oliver; Arcos, Teresa de los; Benedikt, Jan; Keudell, Achim von [RD Plasmas with Complex Interactions, Ruhr-Universität Bochum, Universitätsstr. 150, 44780 Bochum (Germany)

2013-10-15

A beam experiment is presented to study heterogeneous reactions relevant to plasma-surface interactions in reactive sputtering applications. Atom and ion sources are focused onto the sample to expose it to quantified beams of oxygen, nitrogen, hydrogen, noble gas ions, and metal vapor. The heterogeneous surface processes are monitored in situ by means of a quartz crystal microbalance and Fourier transform infrared spectroscopy. Two examples illustrate the capabilities of the particle beam setup: oxidation and nitriding of aluminum as a model of target poisoning during reactive magnetron sputtering, and plasma pre-treatment of polymers (PET, PP)

Study of transient rod extraction failure without RBM in a BWR

International Nuclear Information System (INIS)

Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

2015-09-01

The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

Localized reactive flow in carbonate rocks: Core-flood experiments and network simulations

Science.gov (United States)

Wang, Haoyue; Bernabé, Yves; Mok, Ulrich; Evans, Brian

2016-11-01

We conducted four core-flood experiments on samples of a micritic, reef limestone from Abu Dhabi under conditions of constant flow rate. The pore fluid was water in equilibrium with CO2, which, because of its lowered pH, is chemically reactive with the limestone. Flow rates were between 0.03 and 0.1 mL/min. The difference between up and downstream pore pressures dropped to final values ≪1 MPa over periods of 3-18 h. Scanning electron microscope and microtomography imaging of the starting material showed that the limestone is mostly calcite and lacks connected macroporosity and that the prevailing pores are few microns large. During each experiment, a wormhole formed by localized dissolution, an observation consistent with the decreases in pressure head between the up and downstream reservoirs. Moreover, we numerically modeled the changes in permeability during the experiments. We devised a network approach that separated the pore space into competing subnetworks of pipes. Thus, the problem was framed as a competition of flow of the reactive fluid among the adversary subnetworks. The precondition for localization within certain time is that the leading subnetwork rapidly becomes more transmissible than its competitors. This novel model successfully simulated features of the shape of the wormhole as it grew from few to about 100 µm, matched the pressure history patterns, and yielded the correct order of magnitude of the breakthrough time. Finally, we systematically studied the impact of changing the statistical parameters of the subnetworks. Larger mean radius and spatial correlation of the leading subnetwork led to faster localization.

Guanosine radical reactivity explored by pulse radiolysis coupled with transient electrochemistry.

Science.gov (United States)

Latus, A; Alam, M S; Mostafavi, M; Marignier, J-L; Maisonhaute, E

2015-06-04

We follow the reactivity of a guanosine radical created by a radiolytic electron pulse both by spectroscopic and electrochemical methods. This original approach allows us to demonstrate that there is a competition between oxidation and reduction of these intermediates, an important result to further analyse the degradation or repair pathways of DNA bases.

Transient feedback from fuel motion in metal IFR [Integral Fast Reactor] fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Stanford, G.S.; Regis, J.P.; Bauer, T.H.; Dickerman, C.E.

1990-01-01

Results from hodoscope data analyses are presented for TREAT transient-overpower tests M5 through M7 with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding branch and prefailure elongation of D9-clad ternary (U-Pu-Zr) IFR-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT9-clad binary (U-Zr) FFTF-driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure. 4 refs., 6 figs

Nuclear data propagation with burnup. Impact on SFR reactivity coefficients

International Nuclear Information System (INIS)

Buiron, Laurent; Plisson-Rieunier, Daniele

2017-01-01

For the next generation fast reactor design, the Generation IV International Forum (GIF) defined global objectives in terms of safety improvement, sustainability, waste minimization and non-proliferation. Among the possibilities studied at CEA, Sodium cooled Fast Reactor (SFR) are studied as potential industrial tools for next decade's deployment. Many efforts have been made in the last years to obtain advanced industrial core designs that comply with these goals. Concerning safety issues, particular efforts have been made in order to obtain core designs that can be resilient to accidental transients. The 'safety' level of such new designs is often characterized by their 'natural' behavior under unprotected transients such as loss of flow or hypothetical transient over power. Transient analysis needs several accurate neutronic input data such as reactivity coefficient and kinetic parameters. Beside estimation of the level of 'absolute' values, associated uncertainties have also to be evaluated for the whole set of relevant data. These estimations have to be performed for different core state such as end of cycle core for feedback coefficient. This means that uncertainties have to be obtained not only a fixed time but also have to be propagated all through irradiation. To do so, we need to couple Boltzman and Bateman equations at sensitivities level. The coupling process could be done with the help of the perturbation theory which gives adapted framework suited for deterministic calculation codes. This coupling is currently in progress in ERANOS code system. The actual implementation gives access to estimation of sensitivities for both reactivity coefficients and mass balance. After a brief theoretical description of Boltzman/Bateman coupling capabilities in ERANOS, the study presented in this paper focuses on sensitivity and uncertainties estimation for the main feedback coefficients involved in fast reactor transients: the

RETRAN experience with BWR transients at Yankee Atomic Electric Company

International Nuclear Information System (INIS)

Ansari, A.A.F.; Cronin, J.T.; Slifer, B.C.

1981-01-01

Yankee Atomic Electric Company is actively involved in the development of licensing methods for BWR's. The computer code chosen for analyzing system response under transient conditions is RETRAN. This paper describes the RETRAN model developed for Vermont Yankee, and the results of the RETRAN checkout and qualification that has been achieved at YAEC through comparison of RETRAN predictions to the startup test results performed at the plant as part of the 100% power startup test program. In addition, abnormal operational transients typically analyzed for licensing are also presented

Cerebral blood flow and CO2 reactivity in transient ischemic attacks: comparison between TIAs due to the ICA occlusion and ICA mild stenosis

International Nuclear Information System (INIS)

Tsuda, Y.; Kimura, K.; Yoneda, S.; Etani, H.; Asai, T.; Nakamura, M.; Abe, H.

1983-01-01

Hemispheric mean cerebral blood flow (CBF), together with its CO2 reactivity in response to hyperventilation, was investigated in 18 patients with transient ischemic attacks (TIAs) by intraarterial 133Xe injection method in a subacute-chronic stage of the clinical course. In 8 patients, the lesion responsible for symptoms was regarded as unilateral internal carotid artery (ICA) occlusion, and in 10 patients, it was regarded as unilateral ICA mild stenosis (less than 50% stenosis in diameter). Resting flow values were significantly decreased in the affected hemisphere of TIA due to the ICA occlusion as compared with the unaffected hemisphere of the same patient, regarded as the relative control. It was not decreased in the affected hemisphere of TIA due to the ICA mild stenosis as compared with the control. With respect to the responsiveness of CBF to changes in PaCO2, it was preserved in both TIAs, due to the ICA occlusion and ICA mild stenosis. Vasoparalysis was not observed in either types of TIAs in the subacute-chronic stage. However, in the relationship of blood pressure and CO2 reactivity, expressed as delta CBF(%)/delta PaCO2, pressure-dependent CO2 reactivity as a group was observed with significance in 8 cases of TIA due to the ICA occlusion, while no such relationship was noted in 10 cases of TIA due to the ICA mild stenosis. Moreover, clinical features were different between TIAs due to the ICA occlusion and ICA mild stenosis, i.e., more typical, repeatable TIA (6.3 +/- 3.7 times) with shorter duration (less than 30 minutes) was observed in TIAs due to the ICA mild stenosis, while more prolonged, less repeatable TIA (2.4 +/- 1.4 times) was observed in TIAs due to fixed obstruction of the ICA. From these observations, two different possible mechanisms as to the pathogenesis of TIA might be expected

An Appetitive Experience after Fear Memory Destabilization Attenuates Fear Retention: Involvement GluN2B-NMDA Receptors in the Basolateral Amygdala Complex

Science.gov (United States)

Ferrer Monti, Roque I.; Giachero, Marcelo; Alfei, Joaquín M.; Bueno, Adrián M.; Cuadra, Gabriel; Molina, Victor A.

2016-01-01

It is known that a consolidated memory can return to a labile state and become transiently malleable following reactivation. This instability is followed by a restabilization phase termed reconsolidation. In this work, we explored whether an unrelated appetitive experience (voluntary consumption of diluted sucrose) can affect a contextual fear…

A faster reactor transient analysis methodology for PCs

International Nuclear Information System (INIS)

Ott, K.O.

1991-10-01

The simplified ANL model for LMR transient analysis, in which point kinetics as well as lumped descriptions of the heat transfer equations in all components are applied, is converted from a differential into an integral formulation. All differential balance equations are implicitly solved in terms of convolution integrals. The prompt jump approximation is applied as the strong negative feedback effectively keeps the net reactivity well below prompt critical. After implicit finite differencing of the convolution integrals, the kinetics equation assumes the form of a quadratic equation, the ''quadratic dynamics equation.'' This model forms the basis for GW-BASIC program, LTC, for LMR Transient Calculation program, which can effectively be run on a PC. The GW-BASIC version of the LTC program is described in detail in Volume 2 of this report

Reactive transport modelling of a heating and radiation experiment in the Boom clay (Belgium)

International Nuclear Information System (INIS)

Montenegro, L.; Samper, J.; Delgado, J.

2003-01-01

Most countries around the world consider Deep Geological Repositories (DGR) as the most safe option for the final disposal of high level radioactive waste (HLW). DGR is based on adopting a system of multiple barriers between the HLW and the biosphere. Underground laboratories provide information about the behaviour of these barriers at real conditions. Here we present a reactive transport model for the CERBERUS experiment performed at the HADES underground laboratory at Mol (Belgium) in order to characterize the thermal (T), hydrodynamic (H) and geochemical (G) behaviour of the Boon clay. This experiment is unique because it addresses the combined effect of heat and radiation produced by the storage of HLW in a DGR. Reactive transport models which are solved with CORE, are used to perform quantitative predictions of Boom clay thermo-hydro-geochemical (THG) behaviour. Numerical results indicate that heat and radiation cause a slight oxidation near of the radioactive source, pyrite dissolution, a pH decrease and slight changes in the pore water chemical composition of the Boom clay. (Author) 33 refs

A transient kinetic study of nickel-catalyzed methanation: Final report

International Nuclear Information System (INIS)

Hoost, T.E.; Goodwin, J.G. Jr.

1988-11-01

The results of this study are in two major parts. In Part I the use of steady-state isotopic transients of multiple elements (C, H, and O) under actual methanation reaction conditions has permitted an assessment of the reactivity of water on a Ni powder catalyst. It was concluded based on the addition of isotopic water that oxygen, once reacted to form water, is able to readsorb even where the surface coverage of CO remains high. At the low relative partial pressures of water used, however, there was no effect of added water on the formation of methane. The surface residence time of water determined from isotopic transients contains the residence time on the surface during the primary formation reaction as well as the time spent during readsorption(s). Part II addressed how a catalyst modifier (in this case Cl) affects methanation in CO hydrogenation using steady-state isotopic transient kinetic analysis (SSITKA) of methanation. The results obtained using silica-supported Ru suggest the structural rearrangements induced by the presence of chlorine, rather than selective site blocking or electronic interactions, may be the primary mechanism of chlorine modification of the catalytic properties of supported metals for CO hydrogenation. SSITKA indicated that the decrease in methanation activity with increasing initial Cl concentration was a function of a decrease in the number of reactive surface intermediates (or sites) and not of a change in site activity. 36 refs., 10 figs., 5 tabs

Techniques for computing reactivity changes caused by fuel axial expansion in LMR's

International Nuclear Information System (INIS)

Khalil, H.

1988-01-01

An evaluation is made of the accuracy of methods used to compute reactivity changes caused by axial fuel relocation in fast reactors. Results are presented to demonstrate the validity of assumptions commonly made such as linearity of reactivity with fuel elongation, additivity of local reactivity contributions, and the adequacy of standard perturbation techniques. Accurate prediction of the reactivity loss caused by axial swelling of metallic fuel is shown to require proper representation of the burnup dependence of the expansion reactivity. Some accuracy limitations in the methods used in transient analyses, which are based on the use of fuel worth tables, are identified, and efficient ways to improve accuracy are described. Implementation of these corrections produced expansion reactivity estimates within 5% of higher-order method for a metal-fueled FFTF core representation. 18 refs., 3 figs., 3 tabs

Altered Brain Reactivity to Game Cues After Gaming Experience.

Science.gov (United States)

Ahn, Hyeon Min; Chung, Hwan Jun; Kim, Sang Hee

2015-08-01

Individuals who play Internet games excessively show elevated brain reactivity to game-related cues. This study attempted to test whether this elevated cue reactivity observed in game players is a result of repeated exposure to Internet games. Healthy young adults without a history of excessively playing Internet games were recruited, and they were instructed to play an online Internet game for 2 hours/day for five consecutive weekdays. Two control groups were used: the drama group, which viewed a fantasy TV drama, and the no-exposure group, which received no systematic exposure. All participants performed a cue reactivity task with game, drama, and neutral cues in the brain scanner, both before and after the exposure sessions. The game group showed an increased reactivity to game cues in the right ventrolateral prefrontal cortex (VLPFC). The degree of VLPFC activation increase was positively correlated with the self-reported increase in desire for the game. The drama group showed an increased cue reactivity in response to the presentation of drama cues in the caudate, posterior cingulate, and precuneus. The results indicate that exposure to either Internet games or TV dramas elevates the reactivity to visual cues associated with the particular exposure. The exact elevation patterns, however, appear to differ depending on the type of media experienced. How changes in each of the regions contribute to the progression to pathological craving warrants a future longitudinal study.

Elucidating reactivity regimes in cyclopentane oxidation: Jet stirred reactor experiments, computational chemistry, and kinetic modeling

KAUST Repository

Rachidi, Mariam El; Thion, Sé bastien; Togbé , Casimir; Dayma, Guillaume; Mehl, Marco; Dagaut, Philippe; Pitz, William J.; Zá dor, Judit; Sarathy, Mani

2016-01-01

This study is concerned with the identification and quantification of species generated during the combustion of cyclopentane in a jet stirred reactor (JSR). Experiments were carried out for temperatures between 740 and 1250K, equivalence ratios from 0.5 to 3.0, and at an operating pressure of 10atm. The fuel concentration was kept at 0.1% and the residence time of the fuel/O/N mixture was maintained at 0.7s. The reactant, product, and intermediate species concentration profiles were measured using gas chromatography and Fourier transform infrared spectroscopy. The concentration profiles of cyclopentane indicate inhibition of reactivity between 850-1000K for ϕ = 2.0 and ϕ = 3.0. This behavior is interesting, as it has not been observed previously for other fuel molecules, cyclic or non-cyclic. A kinetic model including both low- and high-temperature reaction pathways was developed and used to simulate the JSR experiments. The pressure-dependent rate coefficients of all relevant reactions lying on the PES of cyclopentyl+O, as well as the C-C and C-H scission reactions of the cyclopentyl radical were calculated at the UCCSD(T)-F12b/cc-pVTZ-F12//M06-2X/6-311++G(d,p) level of theory. The simulations reproduced the unique reactivity trend of cyclopentane and the measured concentration profiles of intermediate and product species. Sensitivity and reaction path analyses indicate that this reactivity trend may be attributed to differences in the reactivity of allyl radical at different conditions, and it is highly sensitive to the C-C/C-H scission branching ratio of the cyclopentyl radical decomposition.

Elucidating reactivity regimes in cyclopentane oxidation: Jet stirred reactor experiments, computational chemistry, and kinetic modeling

KAUST Repository

Rachidi, Mariam El

2016-06-23

This study is concerned with the identification and quantification of species generated during the combustion of cyclopentane in a jet stirred reactor (JSR). Experiments were carried out for temperatures between 740 and 1250K, equivalence ratios from 0.5 to 3.0, and at an operating pressure of 10atm. The fuel concentration was kept at 0.1% and the residence time of the fuel/O/N mixture was maintained at 0.7s. The reactant, product, and intermediate species concentration profiles were measured using gas chromatography and Fourier transform infrared spectroscopy. The concentration profiles of cyclopentane indicate inhibition of reactivity between 850-1000K for ϕ = 2.0 and ϕ = 3.0. This behavior is interesting, as it has not been observed previously for other fuel molecules, cyclic or non-cyclic. A kinetic model including both low- and high-temperature reaction pathways was developed and used to simulate the JSR experiments. The pressure-dependent rate coefficients of all relevant reactions lying on the PES of cyclopentyl+O, as well as the C-C and C-H scission reactions of the cyclopentyl radical were calculated at the UCCSD(T)-F12b/cc-pVTZ-F12//M06-2X/6-311++G(d,p) level of theory. The simulations reproduced the unique reactivity trend of cyclopentane and the measured concentration profiles of intermediate and product species. Sensitivity and reaction path analyses indicate that this reactivity trend may be attributed to differences in the reactivity of allyl radical at different conditions, and it is highly sensitive to the C-C/C-H scission branching ratio of the cyclopentyl radical decomposition.

The VLA Low-band Ionosphere and Transient Experiment (VLITE)

Science.gov (United States)

Clarke, Tracy; Peters, Wendy; Brisken, Walter; Giacintucci, Simona; Kassim, Namir; Polisensky, Emil; Helmboldt, Joseph; Richards, Emily E.; Erickson, Alan; Ray, Paul S.; Kerr, Matthew T.; Deneva, Julia; Coburn, William; Huber, Robert; Long, Jeff

2018-01-01

The VLA Low-band Ionosphere and Transient Experiment (VLITE, http://vlite.nrao.edu/ ) is a commensal low-frequency observing system that has been operational on the National Radio Astronomy Observatory's Karl G. Jansky Very Large Array (VLA) since late 2014. The separate optical paths of the prime-focus sub-GHz dipole feeds and the Cassegrain-focus 1-50 GHz feeds allow both systems to operate simultaneously with independent correlators. The initial 2.5 years of VLITE operation provided real-time correlation of 10 antennas across the 320-384 MHz band with a total observing time approaching 12,000 hours. During the summer of 2017, VLITE was upgraded to a total of 16 antennas (more than doubling the number of baselines) with enhanced correlator capabilities to enable correlation of the on-the-fly observing mode being used for the new NRAO VLA Sky Survey (VLASS).We present an overview of the VLITE system, including highlights of the complexities of a commensal observing program, sparse-array challenges, and scientific capabilities from our science-ready data pipeline. In the longer term, we seek a path to broadband expansion across all VLA antennas to develop a powerful new LOw Band Observatory (LOBO).

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

Biogeochemical Reactive Transport Model of the Redox Zone Experiment of the sp Hard Rock Laboratory in Sweden

International Nuclear Information System (INIS)

Molinero-Huguet, Jorge; Samper-Calvete, F. Javier; Zhang, Guoxiang; Yang, Changbing

2004-01-01

Underground facilities are being operated by several countries around the world for performing research and demonstration of the safety of deep radioactive waste repositories. The ''sp'' Hard Rock Laboratory is one such facility launched and operated by the Swedish Nuclear Fuel and Waste Management Company where various in situ experiments have been performed in fractured granites. One such experiment is the redox zone experiment, which aimed at evaluating the effects of the construction of an access tunnel on the hydrochemical conditions of a fracture zone. Dilution of the initially saline groundwater by fresh recharge water is the dominant process controlling the hydrochemical evolution of most chemical species, except for bicarbonate and sulfate, which unexpectedly increase with time. We present a numerical model of water flow, reactive transport, and microbial processes for the redox zone experiment. This model provides a plausible quantitatively based explanation for the unexpected evolution of bicarbonate and sulfate, reproduces the breakthrough curves of other reactive species, and is consistent with previous hydrogeological and solute transport models

Reactivity worth measurements on the CALIBAN reactor: interpretation of integral experiments for the nuclear data validation

International Nuclear Information System (INIS)

Richard, B.

2012-01-01

The good knowledge of nuclear data, input parameters for the neutron transport calculation codes, is necessary to support the advances of the nuclear industry. The purpose of this work is to bring pertinent information regarding the nuclear data integral validation process. Reactivity worth measurements have been performed on the Caliban reactor, they concern four materials of interest for the nuclear industry: gold, lutetium, plutonium and uranium 238. Experiments which have been conducted in order to improve the characterization of the core are also described and discussed, the latter are necessary to the good interpretation of reactivity worth measurements. The experimental procedures are described with their associated uncertainties, measurements are then compared to numerical results. The methods used in numerical calculations are reported, especially the multigroup cross sections generation for deterministic codes. The modeling of the experiments is presented along with the associated uncertainties. This comparison led to an interpretation concerning the qualification of nuclear data libraries. Discrepancies are reported, discussed and justify the need of such experiments. (author) [fr

Features of Onset and Clinical Course of Reactive Arthritis in Children

Directory of Open Access Journals (Sweden)

I.S. Lebets

2013-09-01

Results. Reactive arthritis of chlamydial etiology is characterized by lesion of large and medium-sized joints of the lower limbs, which is often accompanied by short-term morning stiffness and rapid onset of transient hypomyatrophy. Reiter’s disease may develop rarely. Mycoplasma-induced reactive arthritis is characterized by debut with arthritis of knee, ankle, wrist and small joints of the hand, the development of bursitis and hypomyatrophy. Feature of Ureaplasma arthritis is the formation of bursitis in the heel and tendinitis. Reactive arthritis associated with elevated titers to antistreptolysin O differs with polymorphism of articular syndrome manifestations and, to some extent, of similarity with juvenile rheumatoid arthritis. Unspecified reactive arthritis has a number of the general features with others reactive arthritis and it is characterized by rather benign clinical course, long preservation of joints function and low laboratory activity. Relapse rate of reactive arthritis increases with an increase of duration of illness.

Transient fuel and target performance testing for the HWR-NPR

International Nuclear Information System (INIS)

Jicha, J.J. Jr.

1990-01-01

This paper describes a five year program of fuel target transient performance testing and model development required for the safety assessment of the HWR new production reactor. Technical issues are described, focusing on fuel and target behavior during extremely low probability transients which can lead to fuel melting. Early work on these issues is reviewed. The program to meet remaining needs is described. Three major transient-testing activities are included: in-cell experiments on small samples of irradiated fuel and target, small-scale phenomenological experiments in the ACRR reactor, and limited-integral experiments in the TREAT reactor. A coordinated development of detailed fuel and target behavior models is also described

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

Description and characterization of the ACRR's programmable transient rod withdrawal mode

International Nuclear Information System (INIS)

Boldt, K.R.; Sullivan, W.H.; Kefauver, H.L.

1980-01-01

To satisfy experiment needs for Sandia's Advanced Reactor Safety Program, a programmable Transient Rod Withdrawal (TRW) mode has been developed for the Annular Core Research Reactor (ACRR). The programmable mode is a modification of the existing continuous-withdrawal TRW mode and permits speed and direction changes during the pulse sequence. Basically, a TRW operation is similar to a routine pulse operation except that transient rods are mechanically withdrawn rather than pneumatically fired. Being a pulse-type operation, the TRW mode complies with pulse-mode safety system settings. Control system interlocks prevent the pneumatic firing of rods in the TRW mode. The hardware for the programmable TRW mode includes three ACRR transient rods, the ACRR timer, two rod programmers, a minicomputer and a summing circuit for position indication. Each ACRR transient rod is mechanically driven by a stepping motor (rated torque at 4.24 joules) and is capable of a maximum TRW speed of 26.7 centimeters/ second. The maximum reactivity insertion rate is $2.45/second with a transient rod bank worth of $3.00 and $3.47/second with a bank worth of $4.25, which is expected to be installed soon. The ACRR timer is a multifunctional timer used in all operating modes of the reactor. In the programmable TRW mode, the timer starts the rod programmers and drops regulating rods to terminate the operation. Programmed withdrawal capability is provided by one of two rod programmers (a hardwire-based unit and a microprocessor-based unit). The hardwire unit has eight intervals in which speed, direction and distance are selected by switches on the front panel. The microprocessor-based unit has the capability of 64 intervals in which speed, direction, and distance or time can be specified. Programming this unit is accomplished from the front panel or by inputting data from an HP-9845. minicomputer via a digital I/O interface. Self-test programs in the software provide a continual check of an operating

The Everyday Emotional Experience of Adults with Major Depressive Disorder: Examining Emotional Instability, Inertia, and Reactivity

Science.gov (United States)

Thompson, Renee J.; Mata, Jutta; Jaeggi, Susanne M.; Buschkuehl, Martin; Jonides, John; Gotlib, Ian H.

2013-01-01

Investigators have begun to examine the temporal dynamics of affect in individuals diagnosed with Major Depressive Disorder (MDD), focusing on instability, inertia, and reactivity of emotion. How these dynamics differ between individuals with MDD and healthy controls have not before been examined in a single study. In the present study, 53 adults with MDD and 53 healthy adults carried hand-held electronic devices for approximately seven days and were prompted randomly eight times per day to report their levels of current negative affect (NA), positive affect (PA), and the occurrence of significant events. In terms of NA, compared with healthy controls, depressed participants reported greater instability and greater reactivity to positive events, but comparable levels of inertia and reactivity to negative events. Neither average levels of NA nor NA reactivity to, frequency or intensity of, events accounted for the group difference in instability of NA. In terms of PA, the MDD and control groups did not differ significantly in their instability, inertia, or reactivity to positive or negative events. These findings highlight the importance of emotional instability in MDD, particularly with respect to NA, and contribute to a more nuanced understanding of the everyday emotional experiences of depressed individuals. PMID:22708886

Investigation of natural circulation instability and transients in passively safe novel modular reactor

Science.gov (United States)

Shi, Shanbin

The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

Analysis of reactivity feedback effects of void and temperature in the MAPLE-X10 reactor

International Nuclear Information System (INIS)

Carlson, P.A.; Heeds, W.; Shim, S.Y.; King, S.G.

1992-07-01

The methods used for evaluating the void and temperature reactivity coefficients for the MAPLE-X10 Reactor are described and factors used in estimating their accuracy are discussed. The report presents representative transient analysis results using the CATHENA thermalhydraulics code. The role of the reactivity coefficients and their precision is discussed. The results are reviewed in terms of their safety implications

Fuel-pin cladding transient failure strain criterion

International Nuclear Information System (INIS)

Bard, F.E.; Duncan, D.R.; Hunter, C.W.

1983-01-01

A criterion for cladding failure based on accumulated strain was developed for mixed uranium-plutonium oxide fuel pins and used to interpret the calculated strain results from failed transient fuel pin experiments conducted in the Transient Reactor Test (TREAT) facility. The new STRAIN criterion replaced a stress-based criterion that depends on the DORN parameter and that incorrectly predicted fuel pin failure for transient tested fuel pins. This paper describes the STRAIN criterion and compares its prediction with those of the stress-based criterion

Contributions of primary and secondary biogenic VOC tototal OH reactivity during the CABINEX (Community Atmosphere-Biosphere INteractions Experiments-09 field campaign

Directory of Open Access Journals (Sweden)

S. Kim

2011-08-01

Full Text Available We present OH reactivity measurements using the comparative reactivity method with a branch enclosure technique for four different tree species (red oak, white pine, beech and red maple in the UMBS PROPHET tower footprint during the Community Atmosphere Biosphere INteraction EXperiment (CABINEX field campaign in July of 2009. Proton Transfer Reaction-Mass Spectrometry (PTR-MS was sequentially used as a detector for OH reactivity and BVOC concentrations including isoprene and monoterpenes (MT for enclosure air. Therefore, the measurement dataset contains both measured and calculated OH reactivity from well-known BVOC. The results indicate that isoprene and MT, and in one case a sesquiterpene, can account for the measured OH reactivity. Significant discrepancy between measured OH reactivity and calculated OH reactivity from isoprene and MT is found for the red maple enclosure dataset but it can be reconciled by adding reactivity from emission of a sesquiterpene, α-farnesene, detected by GC-MS. This leads us to conclude that no significant unknown BVOC emission contributed to ambient OH reactivity from these trees at least during the study period. However, this conclusion should be followed up by more comprehensive side-by-side intercomparison between measured and calculated OH reactivity and laboratory experiments with controlled temperature and light environments to verify effects of those essential parameters towards unknown/unmeasured reactive BVOC emissions. This conclusion leads us to explore the contribution towards ambient OH reactivity (the dominant OH sink in this ecosystem oxidation products such as hydroxyacetone, glyoxal, methylglyoxal and C4 and C5-hydroxycarbonyl using recently published isoprene oxidation mechanisms (Mainz Isoprene Mechanism II and Leuven Isoprene Mechanism. Evaluation of conventionally unmeasured first generation oxidation products of isoprene and their possible contribution to ambient missing OH reactivity

Irradiation creep transients in Ni-4 at.% Si

International Nuclear Information System (INIS)

Nagakawa, J.

1983-01-01

In the course of irradiation creep experiments on Ni-4 at.% Si alloy, two types of creep transients were observed on the termination of irradiation. The short term transient was completed within one minute while the long term transient persisted for nearly ten hours. A change in the temperature distribution was excluded from the possible causes, partly because the stress dependence of the observed transient strains was not linear, and partly because the strain increase expected from the temperature change was much smaller than the observed value. Transient behavior of point defects was examined in conjunction with the climb-glide mechanism and the steady-state irradiation creep data. Calculated creep transient due to excess vacancy flux to dislocations was in good agreement with the observed short term transient. The long term transient appears to be a result of dislocation microstructure change. The present results suggest an enhanced irradiation creep under cyclic irradiation conditions which will be encountered in the early generations of fusion reactors. (orig.)

Transient Control of Synchronous Machine Active and Reactive Power in Micro-grid Power Systems

Science.gov (United States)

Weber, Luke G.

There are two main topics associated with this dissertation. The first is to investigate phase-to-neutral fault current magnitude occurring in generators with multiple zero-sequence current sources. The second is to design, model, and tune a linear control system for operating a micro-grid in the event of a separation from the electric power system. In the former case, detailed generator, AC8B excitation system, and four-wire electric power system models are constructed. Where available, manufacturers data is used to validate the generator and exciter models. A gain-delay with frequency droop control is used to model an internal combustion engine and governor. The four wire system is connected through a transformer impedance to an infinite bus. Phase-to-neutral faults are imposed on the system, and fault magnitudes analyzed against three-phase faults to gauge their severity. In the latter case, a balanced three-phase system is assumed. The model structure from the former case - but using data for a different generator - is incorporated with a model for an energy storage device and a net load model to form a micro-grid. The primary control model for the energy storage device has a high level of detail, as does the energy storage device plant model in describing the LC filter and transformer. A gain-delay battery and inverter model is used at the front end. The net load model is intended to be the difference between renewable energy sources and load within a micro-grid system that has separated from the grid. Given the variability of both renewable generation and load, frequency and voltage stability are not guaranteed. This work is an attempt to model components of a proposed micro-grid system at the University of Wisconsin Milwaukee, and design, model, and tune a linear control system for operation in the event of a separation from the electric power system. The control module is responsible for management of frequency and active power, and voltage and reactive

Impact of Adverse Childhood Experiences on Psychotic-Like Symptoms and Stress Reactivity in Daily Life in Nonclinical Young Adults.

Directory of Open Access Journals (Sweden)

Paula Cristóbal-Narváez

Full Text Available There is increasing interest in elucidating the association of different childhood adversities with psychosis-spectrum symptoms as well as the mechanistic processes involved. This study used experience sampling methodology to examine (i associations of a range of childhood adversities with psychosis symptom domains in daily life; (ii whether associations of abuse and neglect with symptoms are consistent across self-report and interview methods of trauma assessment; and (iii the role of different adversities in moderating affective, psychotic-like, and paranoid reactivity to situational and social stressors.A total of 206 nonclinical young adults were administered self-report and interview measures to assess childhood abuse, neglect, bullying, losses, and general traumatic events. Participants received personal digital assistants that signaled them randomly eight times daily for one week to complete questionnaires about current experiences, including symptoms, affect, and stress.Self-reported and interview-based abuse and neglect were associated with psychotic-like and paranoid symptoms, whereas only self-reported neglect was associated with negative-like symptoms. Bullying was associated with psychotic-like symptoms. Losses and general traumatic events were not directly associated with any of the symptom domains. All the childhood adversities were associated with stress reactivity in daily life. Interpersonal adversities (abuse, neglect, bullying, and losses moderated psychotic-like and/or paranoid reactivity to situational and social stressors, whereas general traumatic events moderated psychotic-like reactivity to situational stress. Also, different interpersonal adversities exacerbated psychotic-like and/or paranoid symptoms in response to distinct social stressors.The present study provides a unique examination of how childhood adversities impact the expression of spectrum symptoms in the real world and lends support to the notion that

Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

International Nuclear Information System (INIS)

Yamamoto, Takahisa; Mitachi, Koshi

2004-01-01

This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

Hidden photoinduced reactivity of the blue fluorescent protein mKalama1

DEFF Research Database (Denmark)

Vegh, Russell B.; Bloch, Dmitry A.; Bommarius, Andreas S.

2015-01-01

, is largely unexplored. Here, by using transient absorption spectroscopy spanning the time scale from picoseconds to seconds, we reveal a hidden reactivity of the bright blue-light emitting protein mKalama1 previously thought to be inert. This protein shows no excited-state proton transfer during its...

Investigation of reactivity between SiC and Nb-1Zr in planned irradiation creep experiments

Energy Technology Data Exchange (ETDEWEB)

Lewinsohn, C.A.; Hamilton, M.L.; Jones, R.H.

1997-08-01

Thermodynamic calculations and diffusion couple experiments showed that SiC and Nb-1Zr were reactive at the upper range of temperatures anticipated in the planned irradiation creep experiment. Sputter-deposited aluminum oxide (Al{sub 2}O{sub 3}) was selected as a diffusion barrier coating. Experiments showed that although the coating coarsened at high temperature it was an effective barrier for diffusion of silicon from SiC into Nb-1Zr. Therefore, to avoid detrimental reactions between the SiC composite and the Nb-1Zr pressurized bladder during the planned irradiation creep experiment, a coating of Al{sub 2}O{sub 3} will be required on the Nb-1Zr bladder.

Analysis of reactivity insertion accidents in PWR reactors

International Nuclear Information System (INIS)

Camargo, C.T.M.

1978-06-01

A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO 2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author) [pt

Advanced methods for BWR transient and stability analysis

Energy Technology Data Exchange (ETDEWEB)

Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

2008-07-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

Advanced methods for BWR transient and stability analysis

International Nuclear Information System (INIS)

Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

2008-01-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

Energy Technology Data Exchange (ETDEWEB)

Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

2016-08-01

A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

TRAB, a transient analysis program for BWR. Part 1

International Nuclear Information System (INIS)

Rajamaeki, Markku.

1980-03-01

TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)

Loss-of-flow transient characterization in carbide-fueled LMFBRs

International Nuclear Information System (INIS)

Rothrock, R.B.; Morgan, M.M.; Baars, R.E.; Elson, J.S.; Wray, M.L.

1985-01-01

One of the benefits derived from the use of carbide fuel in advanced Liquid Metal Fast Breeder Reactors (LMFBRs) is a decreased vulnerability to certain accidents. This can be achieved through the combination of advanced fuel performance with the enhanced reactivity feedback effects and passive shutdown cooling systems characteristic of the current 'inherently safe' plant concepts. The calculated core response to an unprotected loss of flow (ULOF) accident has frequently been used as a benchmark test of these designs, and the advantages of a high-conductivity fuel in relation to this type of transient have been noted in previous analyses. To evaluate this benefit in carbide-fueled LMFBRs incorporating representative current plant design features, limited calculations have been made of a ULOF transient in a small ('modular') carbide-fueled LMFBR

Excitation of neutron flux waves in reactor core transients

International Nuclear Information System (INIS)

Carew, J.F.; Neogy, P.

1983-01-01

An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

Comprehensive Reactive Power Support of DFIG Adapted to Different Depth of Voltage Sags

Directory of Open Access Journals (Sweden)

Yangwu Shen

2017-06-01

Full Text Available The low voltage ride-through (LVRT capability of the doubly-fed induction generator (DFIG significantly impacts upon the integration of wind power into the power grid. This paper develops a novel comprehensive control strategy to enhance the LVRT and reactive power support capacities of the DFIG by installing the energy storage system (ESS. The ESS is connected to the DC-link capacitor of the DFIG and used to regulate the DC-link voltage during normal or fault operations. The unbalanced power between the captured wind power and the power injected to the grid during the transient process is absorbed or compensated by the ESS. The rotor-side converter (RSC is used to control the maximum power production and the grid-side converter (GSC is used to control the reactive power before participating in the voltage support. When the supply voltage continues to drop, the rotor speed is increased by controlling the RSC to realize the LVRT capability and help the GSC further enhance the reactive power support capability. The capacity of the GSC is dedicated to injecting the reactive power to the grid. An auxiliary transient pitch angle controller is proposed to protect the generator’s over speed. Both RSC and GSC act as reactive power sources to further enhance the voltage support capability with serious voltage sags. Simulations based on a single-machine infinite-bus power system verify the effectiveness of the developed comprehensive control strategy.

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.

Study on the reactive transient α-λ3-iodanyl-acetophenone complex in the iodine(III)/PhI(I) catalytic cycle of iodobenzene-catalyzed α-acetoxylation reaction of acetophenone by electrospray ionization tandem mass spectrometry.

Science.gov (United States)

Wang, Hao-Yang; Zhou, Juan; Guo, Yin-Long

2012-03-30

Hypervalent iodine compounds are important and widely used oxidants in organic chemistry. In 2005, Ochiai reported the PhI-catalyzed α-acetoxylation reaction of acetophenone by the oxidation of PhI with m-chloroperbenzoic acid (m-CPBA) in acetic acid. However, until now, the most critical reactive α-λ(3)-iodine alkyl acetophenone intermediate (3) had not been isolated or directly detected. Electrospray ionization tandem mass spectrometry (ESI-MS/MS) was used to intercept and characterize the transient reactive α-λ(3)-iodine alkyl acetophenone intermediate in the reaction solution. The trivalent iodine species was detected when PhI and m-CPBA in acetic acid were mixed, which indicated the facile oxidation of a catalytic amount of PhI(I) to the iodine(III) species by m-CPBA. Most importantly, 3·H(+) was observed at m/z 383 from the reaction solution and this ion gave the protonated α-acetoxylation product 4·H(+) at m/z 179 in MS/MS by an intramolecular reductive elimination of PhI. These ESI-MS/MS studies showed the existence of the reactive α-λ(3)-iodine alkyl acetophenone intermediate 3 in the catalytic cycle. Moreover, the gas-phase reactivity of 3·H(+) was consistent with the proposed solution-phase reactivity of the α-λ(3)-iodine alkyl acetophenone intermediate 3, thus confirming the reaction mechanism proposed by Ochiai. Copyright © 2012 John Wiley & Sons, Ltd.

Attenuation of pyrite oxidation with a fly ash pre-barrier: Reactive transport modelling of column experiments

Energy Technology Data Exchange (ETDEWEB)

Perez-Lopez, R.; Cama, J.; Nieto, J.M.; Ayora, C.; Saaltink, M.W. [University of Huelva, Huelva (Spain). Dept. of Geology

2009-09-15

Conventional permeable reactive barriers (PRBs) for passive treatment of groundwater contaminated by acid mine drainage (AMD) use limestone as reactive material that neutralizes water acidity. However, the limestone-alkalinity potential ceases as inevitable precipitation of secondary metal-phases on grain surfaces occurs, limiting its efficiency. In the present study, fly ash derived from coal combustion is investigated as an alternative alkalinity generating material for the passive treatment of AMD using solution-saturated column experiments. Unlike conventional systems, the utilization of fly ash in a pre-barrier to intercept the non-polluted recharge water before this water reacts with pyrite-rich wastes is proposed. Chemical variation in the columns was interpreted with the reactive transport code RETRASO. In parallel, kinetics of fly ash dissolution at alkaline pH were studied using flow-through experiments and incorporated into the model. In a saturated column filled solely with pyritic sludge-quartz sand (1: 10), oxidation took place at acidic conditions (pH 3.7). According to SO{sub 4}{sup 2-} release and pH, pyrite dissolution occurred favourably in the solution-saturated porous medium until dissolved O{sub 2} was totally consumed. In a second saturated column, pyrite oxidation took place at alkaline conditions (pH 10.45) as acidity was neutralized by fly ash dissolution in a previous level. At this pH Fe release from pyrite dissolution was immediately depleted as Fe-oxy(hydroxide) phases that precipitated on the pyrite grains, forming Fe-coatings (microencapsulation). With time, pyrite microencapsulation inhibited oxidation in practically 97% of the pyritic sludge. Rapid pyrite-surface passivation decreased its reactivity, preventing AMD production in the relatively short term.

Summarizing evaluation of the results of in-pile experiments for the transient fission gas release under accidental conditions of fast breeders

International Nuclear Information System (INIS)

Fischer, E.A.; Vaeth, L.

1989-04-01

The transient fission gas behaviour and the fission gas induced fuel motion were studied in in-pile experiments in different countries, under conditions typical for hypothetical accidents. This report summarizes first the different experiment series and the main results. Then, a comparative evaluation is given, which provides a basis for the choice of the fission gas parameters in the accident code SAS3D

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

Focused cognitive control in dishonesty: Evidence for predominantly transient conflict adaptation.

Science.gov (United States)

Foerster, Anna; Pfister, Roland; Schmidts, Constantin; Dignath, David; Wirth, Robert; Kunde, Wilfried

2018-04-01

Giving a dishonest response to a question entails cognitive conflict due to an initial activation of the truthful response. Following conflict monitoring theory, dishonest responding could therefore elicit transient and sustained control adaptation processes to mitigate such conflict, and the current experiments take on the scope and specificity of such conflict adaptation in dishonesty. Transient adaptation reduces differences between honest and dishonest responding following a recent dishonest response. Sustained adaptation has a similar behavioral signature but is driven by the overall frequency of dishonest responding. Both types of adaptation to recent and frequent dishonest responses have been separately documented, leaving open whether control processes in dishonest responding can flexibly adapt to transient and sustained conflict signals of dishonest and other actions. This was the goal of the present experiments which studied (dis)honest responding to autobiographical yes/no questions. Experiment 1 showed robust transient adaptation to recent dishonest responses whereas sustained control adaptation failed to exert an influence on behavior. It further revealed that transient effects may create a spurious impression of sustained adaptation in typical experimental settings. Experiments 2 and 3 examined whether dishonest responding can profit from transient and sustained adaption processes triggered by other behavioral conflicts. This was clearly not the case: Dishonest responding adapted markedly to recent (dis)honest responses but not to any context of other conflicts. These findings indicate that control adaptation in dishonest responding is strong but surprisingly focused and they point to a potential trade-off between transient and sustained adaptation. (PsycINFO Database Record (c) 2018 APA, all rights reserved).

Study of the initiation of subcooled boiling during power transients

International Nuclear Information System (INIS)

VanVleet, R.J.

1985-01-01

An experimental investigation of boiling initiation during power transients has been conducted for horizontal-cylinder heating elements in degassed distilled water. Platinum elements, 0.127 and 0.250 mm in diameter, were internally heated electrically at a controlled superficial heat flux (power applied divided by surface area) increasing linearly with time at rates of 0.035 and 0.35 MW/m 2 s and corresponding test durations of 20 and 2 seconds. Tests were carried out at saturation temperatures from 100 to 195 0 C with bulk fluid subcooling from 0 to 30 K. During the course of a power transient, element temperature and superficial heat flux were measured electrically and the boiling initiation time was determined optically. It was found that the conditions for boiling initiation depended strongly on the pressure-temperature history of the heating element and surround fluid prior to the transient. Boiling initiation times were found to agree qualitatively with predictions of a model based on the contact-angle hysteresis concept. Brief prepressurization prior to a transient was found to increase dramatically the temperature and heat flux required for boiling initiation because of deactivation of boiling initiation sites. However, sites were re-activated during the transient and, in subsequent tests without prepressurization, no elevation in boiling initiation conditions was observed and results were in quantitative agreement with predictions of the model

Development of a transient criticality evaluation method

International Nuclear Information System (INIS)

Pain, C.C.; Eaton, M.D.; Miles, B.; Ziver, A.K.; Gomes, J.L.M.A.; Umpleby, A.P.; Piggott, M.D.; Goddard, A.J.H.; Oliveira, C.R.E. de

2005-01-01

In developing a transient criticality evaluation method we model, in full spatial/temporal detail, the neutron fluxes and consequent power and the evolving material properties - their flows, energies, phase changes etc. These methods are embodied in the generic method FETCH code which is based as far as possible on basic principles and is capable of use in exploring safety-related situations somewhat beyond the range of experiment. FETCH is a general geometry code capable of addressing a range of criticality issues in fissile materials. The code embodies both transient radiation transport and transient fluid dynamics. Work on powders, granular materials, porous media and solutions is reviewed. The capability for modelling transient criticality for chemical plant, waste matrices and advanced reactors is also outlined. (author)

Transient survivability of LMR oxide fuel pins

International Nuclear Information System (INIS)

Weber, E.T.; Pitner, A.L.; Bard, F.E.; Culley, G.E.; Hunter, C.W.

1986-01-01

Fuel pin integrity during transient events must be assessed for both the core design and safety analysis phases of a reactor project. A significant increase in the experience related to limits of integrity for oxide fuel pins in transient overpower events has been realized from testing of fuel pins irradiated in FFTF and PFR. Fourteen FFTF irradiated fuel pins were tested in TREAT, representing a range of burnups, overpower ramp rates and maximum overpower conditions. Results of these tests along with similar testing in the PFR/TREAT program, provide a demonstration of significant safety margins for oxide fuel pins. Useful information applied in analytical extrapolation of fuel pin test data have been developed from laboratory transient tests on irradiated fuel cladding (FCTT) and on unirradiated fuel pellet deformation. These refinements in oxide fuel transient performance are being applied in assessment of transient capabilities of long lifetime fuel designs using ferritic cladding

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

Pressurizer and steam-generator behavior under PWR transient conditions

International Nuclear Information System (INIS)

Wahba, A.B.; Berta, V.T.; Pointner, W.

1983-01-01

Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

Reactivity costs in MARIA reactor

International Nuclear Information System (INIS)

Marcinkowska, Zuzanna E.; Pytel, Krzysztof M.; Frydrysiak, Andrzej

2017-01-01

Highlights: • The methodology for calculating consumed fuel cost of excess reactivity is proposed. • Correlation between time integral of the core excess reactivity and released energy. • Reactivity price gives number of fuel elements required for given excess reactivity. - Abstract: For the reactor operation at high power level and carrying out experiments and irradiations the major cost of reactor operation is the expense of nuclear fuel. In this paper the methodology for calculating consumed fuel cost-relatedness of excess reactivity is proposed. Reactivity costs have been determined on the basis of operating data. A number of examples of calculating the reactivity costs for processes such as: strong absorbing material irradiation, molybdenium-99 production, beryllium matrix poisoning and increased moderator temperature illustrates proposed method.

The influence of reactive current on wind farm LVRT behavior

Energy Technology Data Exchange (ETDEWEB)

Li, Qing; Zhang, Mei; He, Jing; Qin, Shi-yao [China Electric Power Research Institute, Beijing (China)

2012-07-01

The Low voltage ride through (LVRT) capability of the whole wind farm is required in Chinese grid code published in 2011. In order to analyze the influence of reactive current on wind farm during grid fault, a 100 MW wind farm was simulated with the wind turbines which have been tested. Based on the validated wind turbine model, the wind farm was detailed modelled in DigSILENT/PowerFactory. The model of wind turbines, transformers, feeders, main transformers, static var compensator, and transmission lines was considered in the simulation. Under the weak and strong grid conditions, the wind farm was simulated with different wind turbine reactive current behavior during grid fault, respectively. The voltage distribution, active and reactive power transient behavior at the point of interconnection was analyzed. The results show that wind farm LVRT behavior is related to reactive current and LVRT capability of wind turbine, wind farm electrical structure and grid conditions. And it is very important for wind turbine to have a flexible dynamic reactive current control capability. (orig.)

The economic impact of reactor transients

International Nuclear Information System (INIS)

Rossin, A.D.; Vine, G.L.

1984-01-01

This chapter discusses the cost estimation of transients and the causal relationship between transients and accidents. It is suggested that the calculation of the actual cost of a transient that has occurred is impossible without computerized records. Six months of operating experience reports, based on a survey of pressurized water reactors (PWRs) and boiling water reactors (BWRs) conducted by the Nuclear Safety Analysis Center (NSAC), are analyzed. The significant costs of a reactor transient are the repair costs resulting from severe damage to plant equipment, the cost of scrams (the actions the system is designed to take to avoid safety risks), US NRC fines, negative publicity, utility rates and revenues. It is estimated that the Three Mile Island-2 accident cost the US over $100 billion in nuclear plant delays and cancellations, more expensive fuel, oil imports, backfits, bureaucratic, administrative and legal costs, and lost productivity

Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, S.; Ponsard, B.; Koonen, E.

2007-01-01

The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

Fast Transient And Spatially Non-Homogenous Accident Analysis Of Two-Dimensional Cylindrical Nuclear Reactor

International Nuclear Information System (INIS)

Yulianti, Yanti; Su'ud, Zaki; Waris, Abdul; Khotimah, S. N.; Shafii, M. Ali

2010-01-01

The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.

Generalization of Wilemski-Fixman-Weiss decoupling approximation to the case involving multiple sinks of different sizes, shapes, and reactivities.

Science.gov (United States)

Uhm, Jesik; Lee, Jinuk; Eun, Changsun; Lee, Sangyoub

2006-08-07

We generalize the Wilemski-Fixman-Weiss decoupling approximation to calculate the transient rate of absorption of point particles into multiple sinks of different sizes, shapes, and reactivities. As an application we consider the case involving two spherical sinks. We obtain a Laplace-transform expression for the transient rate that is in excellent agreement with computer simulations. The long-time steady-state rate has a relatively simple expression, which clearly shows the dependence on the diffusion constant of the particles and on the sizes and reactivities of sinks, and its numerical result is in good agreement with the known exact result that is given in terms of recursion relations.

Adaptive Immunity to Leukemia Is Inhibited by Cross-Reactive Induced Regulatory T Cells.

Science.gov (United States)

Manlove, Luke S; Berquam-Vrieze, Katherine E; Pauken, Kristen E; Williams, Richard T; Jenkins, Marc K; Farrar, Michael A

2015-10-15

BCR-ABL(+) acute lymphoblastic leukemia patients have transient responses to current therapies. However, the fusion of BCR to ABL generates a potential leukemia-specific Ag that could be a target for immunotherapy. We demonstrate that the immune system can limit BCR-ABL(+) leukemia progression although ultimately this immune response fails. To address how BCR-ABL(+) leukemia escapes immune surveillance, we developed a peptide: MHC class II tetramer that labels endogenous BCR-ABL-specific CD4(+) T cells. Naive mice harbored a small population of BCR-ABL-specific T cells that proliferated modestly upon immunization. The small number of naive BCR-ABL-specific T cells was due to negative selection in the thymus, which depleted BCR-ABL-specific T cells. Consistent with this observation, we saw that BCR-ABL-specific T cells were cross-reactive with an endogenous peptide derived from ABL. Despite this cross-reactivity, the remaining population of BCR-ABL reactive T cells proliferated upon immunization with the BCR-ABL fusion peptide and adjuvant. In response to BCR-ABL(+) leukemia, BCR-ABL-specific T cells proliferated and converted into regulatory T (Treg) cells, a process that was dependent on cross-reactivity with self-antigen, TGF-β1, and MHC class II Ag presentation by leukemic cells. Treg cells were critical for leukemia progression in C57BL/6 mice, as transient Treg cell ablation led to extended survival of leukemic mice. Thus, BCR-ABL(+) leukemia actively suppresses antileukemia immune responses by converting cross-reactive leukemia-specific T cells into Treg cells. Copyright © 2015 by The American Association of Immunologists, Inc.

TRACE/PARCS modelling of rips trip transients for Lungmen ABWR

Energy Technology Data Exchange (ETDEWEB)

Chang, C. Y. [Inst. of Nuclear Engineering and Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Lin, H. T.; Wang, J. R. [Inst. of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Shih, C. [Inst. of Nuclear Engineering and Science, Dept. of Engineering and System Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China)

2012-07-01

The objectives of this study are to examine the performances of the steady-state results calculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3 code, as well as to use the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen TRACE/PARCS model. In this study, three power generation methods in TRACE were utilized to analyze the three reactor internal pumps (RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general, the comparisons show that the transient responses of key system parameters agree well with the FSAR results, including core power, core inlet flow, reactivity, etc. Further studies will be performed in the future using Lungmen TRACE/PARCS model. After the commercial operation of Lungmen nuclear power plant, TRACE/PARCS model will be verified. (authors)

[VDRL and FTA-ABS reactivity in cerebrospinal fluid: our experience].

Science.gov (United States)

García-Rodríguez, J A; Martín-Sánchez, A M; Canut, A; García-García, L; Cacho, J

1990-01-01

The reactivity of 194 samples of CSF against VDRL and FTA-ABS was studied in patients attending the Clinical Hospital in Salamanca over a five years period. This laboratory was asked to rule out an etiology of syphilis. Twelve samples of CSF proved to be reactive (6.2%) against VDRL and/or FTA-ABS. Seven of these corresponded to six adults diagnosed as suffering from neurosyphilis and one to an infant with early congenital syphilis without neurological alterations; these had in common the presence of active syphilis and a reactive FTA-ABS in serum. In the CSF of the six cases of neurosyphilis, VDRL was reactive in two patients (33.3%) and FTA-ABS in five (83.3%). One minimally reactive VDRL and four FTA-ABS were detected in the remaining five patients, with no known previous history of syphilis, that were suffering from different neurological alterations and that had a nonreactive FTA-ABS in serum. The results obtained in this study point to inappropriate use in CSF of VDRL and FTA-ABS to exclude neurosyphilis in our hospital since only 3.6% of the CSF studied corresponded to patients diagnosed as suffering from neurosyphilis and also to the need for improving the criteria for patient selection.

Weigle Reactivation in Acinetobacter Calcoaceticus

DEFF Research Database (Denmark)

Berenstein, Dvora

1982-01-01

phage and host survivals of about 5 times 10-6 and 1 times 10-1, respectively. Intracellular development of W-reactivated P78 was followed by one-step growth experiments. Conditions which allowed maximal W-reactivation also extended the period of phage production and yielded a somewhat reduced burst......Weigle (W)-reactivation was demonstrated in Acinetobacter calcoaceticus for the UV-irra-diated lysogenic phage P78. The reactivation factor (survival of irradiated phage on irradiated bacteria/ survival on unirradiated bacteria) reached a maximum value of 20. This was obtained at UV-doses giving...

Reactive sites influence in PMMA oligomers reactivity: a DFT study

Science.gov (United States)

Paz, C. V.; Vásquez, S. R.; Flores, N.; García, L.; Rico, J. L.

2018-01-01

In this work, we present a theoretical study of methyl methacrylate (MMA) living anionic polymerization. The study was addressed to understanding two important experimental observations made for Michael Szwarc in 1956. The unexpected effect of reactive sites concentration in the propagation rate, and the self-killer behavior of MMA (deactivating of living anionic polymerization). The theoretical calculations were performed by density functional theory (DFT) to obtain the frontier molecular orbitals values. These values were used to calculate and analyze the chemical interaction descriptors in DFT-Koopmans’ theorem. As a result, it was observed that the longest chain-length species (related with low concentration of reactive sites) exhibit the highest reactivity (behavior associated with the increase of the propagation rate). The improvement in this reactivity was attributed to the crosslinking produced in the polymethyl methacrylate chains. Meanwhile, the self-killer behavior was associated with the intermolecular forces present in the reactive sites. This behavior was associated to an obstruction in solvation, since the active sites remained active through all propagation species. The theoretical results were in good agreement with the Szwarc experiments.

Wind Power Impact to Transient and Voltage Stability of the Power System in Eastern Denmark

DEFF Research Database (Denmark)

Rasmussen, Joana; Jørgensen, Preben; Palsson, Magni Thor

2005-01-01

Voltage stability, transient stability and reactive power compensation are extremely important issues for largescale integration of wind power in areas distant from the main transmission system in Eastern Denmark. This paper describes the application of a dynamic wind farm model in simulation...... studies for assessments of a large wind power penetration. The simulation results reveal problems with voltage stability due to the characteristic of wind turbine generation as well as the inability of the power system to meet the reactive power demand. Furthermore, the established model is applied...

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

Energy Technology Data Exchange (ETDEWEB)

Bruzzone, P. [ed.] [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

2001-01-01

Two Nb{sub 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{sub 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{sup 2}). A transverse pulsed field is applied with {delta}B up to 2.7 T, field rate up to 180 T/s and field integral up to 530 T{sup 2}/s. In the dc test, a good agreement is found between the I{sub c} and the T{sub 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, n{tau}, 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 n{tau} at lower field rate. The calorimetric loss estimation at the fast field transient (f=15 Hz) indicates n{tau} {approx_equal} 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{sup 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

PASP Plus Transient Pulse Monitor (TPM) - Data Analysis and Interpretation Report

National Research Council Canada - National Science Library

Adamo, Richard

1996-01-01

The Transient Pulse Monitor (TPM), part of the PASP Plus experiment aboard the APEX spacecraft, is designed to detect and characterize electromagnetic transient signals produced by electrostatic discharges on the solar array test modules...

Transient Analysis of a Gas-cooled Fast Reactor for Single Control Assembly Withdrawal

International Nuclear Information System (INIS)

Choi, Hangbok

2014-01-01

The Energy Multiplier Module (EMZ) system response has been evaluated for control assembly withdrawal transients. Currently the EM2 core is equipped with six cylindrical drum-type control assemblies in the reflector zone for excess reactivity control and power maneuvering during the operating core life. This study investigates the system response to the control assembly withdrawal accident with various rotational speeds and reactivity worth to determine feasible control assembly design requirements from the physics viewpoint. The simulations have been conducted for single control assembly withdrawal transients without scram by a gas-cooled reactor plant simulator, which is based on a simplified plant nodal model, including the point reactor kinetics, single channel core thermal-fluid model, and a turbo-machinery performance model. Simulations were conducted for the middle-of- cycle core, when the excess reactivity of the core is the highest. Control assembly withdrawal times were varied from 1 (runaway) to 180 sec and reactivity worth was varied from 100 to 400 pcm. For a single control assembly withdrawal, the simulation has shown that the peak fuel temperature is expected to be ~1820°C when the assembly worth is 200 pcm and the runaway time is 1 sec per 180 degree rotation. The peak temperature could be reduced to ~1780°C if the assembly is rotated out in a moderate speed such as 1 degree/sec. These peak temperatures give a thermal margin of 22 to 24% to the melting point of uranium carbide fuel. The results also indicate that the current design with a single control assembly worth of 314 pcm may need adjustments in the future design. (author)

Estimating temperature reactivity coefficients by experimental procedures combined with isothermal temperature coefficient measurements and dynamic identification

International Nuclear Information System (INIS)

Tsuji, Masashi; Aoki, Yukinori; Shimazu, Yoichiro; Yamasaki, Masatoshi; Hanayama, Yasushi

2006-01-01

A method to evaluate the moderator coefficient (MTC) and the Doppler coefficient through experimental procedures performed during reactor physics tests of PWR power plants is proposed. This method combines isothermal temperature coefficient (ITC) measurement experiments and reactor power transient experiments at low power conditions for dynamic identification. In the dynamic identification, either one of temperature coefficients can be determined in such a way that frequency response characteristics of the reactivity change observed by a digital reactivity meter is reproduced from measured data of neutron count rate and the average coolant temperature. The other unknown coefficient can also be determined by subtracting the coefficient obtained from the dynamic identification from ITC. As the proposed method can directly estimate the Doppler coefficient, the applicability of the conventional core design codes to predict the Doppler coefficient can be verified for new types of fuels such as mixed oxide fuels. The digital simulation study was carried out to show the feasibility of the proposed method. The numerical analysis showed that the MTC and the Doppler coefficient can be estimated accurately and even if there are uncertainties in the parameters of the reactor kinetics model, the accuracies of the estimated values are not seriously impaired. (author)

Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT Facility

International Nuclear Information System (INIS)

Varacalle, D.J.; Giri, A.M.; Koizumi, Y.; Koske, J.E.

1983-01-01

An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAPS/MOD1 computer code showed good agreement with the experimental data

Transient waves in visco-elastic media

CERN Document Server

Ricker, Norman

1977-01-01

Developments in Solid Earth Geophysics 10: Transient Waves in Visco-Elastic Media deals with the propagation of transient elastic disturbances in visco-elastic media. More specifically, it explores the visco-elastic behavior of a medium, whether gaseous, liquid, or solid, for very-small-amplitude disturbances. This volume provides a historical overview of the theory of the propagation of elastic waves in solid bodies, along with seismic prospecting and the nature of seismograms. It also discusses the seismic experiments, the behavior of waves propagated in accordance with the Stokes wave

Monte Carlo simulation of nonlinear reactive contaminant transport in unsaturated porous media

International Nuclear Information System (INIS)

Giacobbo, F.; Patelli, E.

2007-01-01

In the current proposed solutions of radioactive waste repositories, the protective function against the radionuclide water-driven transport back to the biosphere is to be provided by an integrated system of engineered and natural geologic barriers. The occurrence of several nonlinear interactions during the radionuclide migration process may render burdensome the classical analytical-numerical approaches. Moreover, the heterogeneity of the barriers' media forces approximations to the classical analytical-numerical models, thus reducing their fidelity to reality. In an attempt to overcome these difficulties, in the present paper we adopt a Monte Carlo simulation approach, previously developed on the basis of the Kolmogorov-Dmitriev theory of branching stochastic processes. The approach is here extended for describing transport through unsaturated porous media under transient flow conditions and in presence of nonlinear interchange phenomena between the liquid and solid phases. This generalization entails the determination of the functional dependence of the parameters of the proposed transport model from the water content and from the contaminant concentration, which change in space and time during the water infiltration process. The corresponding Monte Carlo simulation approach is verified with respect to a case of nonreactive transport under transient unsaturated flow and to a case of nonlinear reactive transport under stationary saturated flow. Numerical applications regarding linear and nonlinear reactive transport under transient unsaturated flow are reported

Oxidation of volatile organic compound vapours by potassium permanganate in a horizontal permeable reactive barrier under unsaturated conditions: experiments and modeling

NARCIS (Netherlands)

Ghareh Mahmoodlu, Mojtaba|info:eu-repo/dai/nl/357287746

2014-01-01

In this research we evaluated the potential of using solid potassium permanganate to create a horizontal permeable reactive barrier (HPRB) for oxidizing VOC vapours in the unsaturated zone. We have performed batch experiments, short column, and long column experiments, and have fully analyzed the

Reactivity feedback models for SSC-K

Energy Technology Data Exchange (ETDEWEB)

Han, Do Hee; Kwon, Young Min; Kim, Kyung Du; Chang, Won Pyo [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

Safety of KALIMER is assured by the inherent safety of the core and passive safety of the safety-related systems. For the safety analysis of a new reactor design such as KALIMER, analysis models, which are consistent with the design, have to be developed for a plant-wide transient and safety analysis code. Efforts for the development of reactivity feedback models for SSC-K, which is now being developed for the safety analysis of KALIMER, is described in this report. Models for Doppler, sodium density/void, fuel axial expansion, core radial expansion, and CRDL expansion have been developed. Test runs have been performed for the unprotected accident for the verification of the models. Use of KALIMER reactivity coefficients and future development of models for GEM and PSDRS would make it possible to analyze the response of KALIMER under TOP as well as LOF and LOHS accident conditions using SSC-K. (author). 5 refs., 64 figs., 2 tabs.

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Variably Saturated Flow and Multicomponent Biogeochemical Reactive Transport Modeling of a Uranium Bioremediation Field Experiment

International Nuclear Information System (INIS)

Yabusaki, Steven B.; Fang, Yilin; Williams, Kenneth H.; Murray, Christopher J.; Ward, Anderson L.; Dayvault, Richard; Waichler, Scott R.; Newcomer, Darrell R.; Spane, Frank A.; Long, Philip E.

2011-01-01

Field experiments at a former uranium mill tailings site have identified the potential for stimulating indigenous bacteria to catalyze the conversion of aqueous uranium in the +6 oxidation state to immobile solid-associated uranium in the +4 oxidation state. This effectively removes uranium from solution resulting in groundwater concentrations below actionable standards. Three-dimensional, coupled variably-saturated flow and biogeochemical reactive transport modeling of a 2008 in situ uranium bioremediation field experiment is used to better understand the interplay of transport rates and biogeochemical reaction rates that determine the location and magnitude of key reaction products. A comprehensive reaction network, developed largely through previous 1-D modeling studies, was used to simulate the impacts on uranium behavior of pulsed acetate amendment, seasonal water table variation, spatially-variable physical (hydraulic conductivity, porosity) and geochemical (reactive surface area) material properties. A principal challenge is the mechanistic representation of biologically-mediated terminal electron acceptor process (TEAP) reactions whose products significantly alter geochemical controls on uranium mobility through increases in pH, alkalinity, exchangeable cations, and highly reactive reduction products. In general, these simulations of the 2008 Big Rusty acetate biostimulation field experiment in Rifle, Colorado confirmed previously identified behaviors including (1) initial dominance by iron reducing bacteria that concomitantly reduce aqueous U(VI), (2) sulfate reducing bacteria that become dominant after ∼30 days and outcompete iron reducers for the acetate electron donor, (3) continuing iron-reducer activity and U(VI) bioreduction during dominantly sulfate reducing conditions, and (4) lower apparent U(VI) removal from groundwater during dominantly sulfate reducing conditions. New knowledge on simultaneously active metal and sulfate reducers has been

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)

Acoustic Transient Source Localization From an Aerostat

National Research Council Canada - National Science Library

Scanlon, Michael; Reiff, Christian; Noble, John

2006-01-01

The Army Research Laboratory (ARL) has conducted experiments using acoustic sensor arrays suspended below tethered aerostats to detect and localize transient signals from mortars, artillery and small arms fire...

Feedback control systems for non-linear simulation of operational transients in LMFBRs

International Nuclear Information System (INIS)

Khatib-Rahbar, M.; Agrawal, A.K.; Srinivasan, E.S.

1979-01-01

Feedback control systems for non-linear simulation of operational transients in LMFBRs are developed. The models include (1) the reactor power control and rod drive mechanism, (2) sodium flow control and pump drive system, (3) steam generator flow control and valve actuator dynamics, and (4) the supervisory control. These models have been incorporated into the SSC code using a flexible approach, in order to accommodate some design dependent variations. The impact of system nonlinearity on the control dynamics is shown to be significant for severe perturbations. Representative result for a 10 cent and 25 cent step insertion of reactivity and a 10% ramp change in load in 40 seconds demonstrate the suitability of this model for study of operational transients without scram in LMFBRs

Unified Behavior Framework for Reactive Robot Control in Real-Time Systems

Science.gov (United States)

2007-03-01

maintain coherent operation in concurrent programs by employing standard communication and synchronization patterns. Some typical ones are: semaphores ...through the semaphore . Signals, whether persistent or transient, are used to communicate between threads as a means of synchronizing their progress...tasks to be decomposed into collections of low-level primitive behaviors, Figure 2.b. This approach takes on the self- contradictory term, reactive

Quadratic reactivity fuel cycle model

International Nuclear Information System (INIS)

Lewins, J.D.

1985-01-01

For educational purposes it is highly desirable to provide simple yet realistic models for fuel cycle and fuel economy. In particular, a lumped model without recourse to detailed spatial calculations would be very helpful in providing the student with a proper understanding of the purposes of fuel cycle calculations. A teaching model for fuel cycle studies based on a lumped model assuming the summability of partial reactivities with a linear dependence of reactivity usefully illustrates fuel utilization concepts. The linear burnup model does not satisfactorily represent natural enrichment reactors. A better model, showing the trend of initial plutonium production before subsequent fuel burnup and fission product generation, is a quadratic fit. The study of M-batch cycles, reloading 1/Mth of the core at end of cycle, is now complicated by nonlinear equations. A complete account of the asymptotic cycle for any order of M-batch refueling can be given and compared with the linear model. A complete account of the transient cycle can be obtained readily in the two-batch model and this exact solution would be useful in verifying numerical marching models. It is convenient to treat the parabolic fit rho = 1 - tau 2 as a special case of the general quadratic fit rho = 1 - C/sub tau/ - (1 - C)tau 2 in suitably normalized reactivity and cycle time units. The parabolic results are given in this paper

Reactors Dynamic analysis Due to Reactivity of The RSG-Gas at One Line Cooling Mode

International Nuclear Information System (INIS)

Hastuti, Endiah Puji

2003-01-01

In the frame of minimizing the operation-cost, operation mode using one line cooling system is being evaluated. Maximum reactor power has been determined and steady state and LOFA transient analysis have also been done. To complete those analyses, the reactivity analysis was done by means of a core dynamic and thermal hydraulic code, PARET-ANL. Accident simulation was done. by a ramp reactivity accident due to control rod withdrawal. Reactivity analysis was carried out at two power range i.e. low and high power level, by imposing one line mode reactor protection limits. The results show that technically, the RSG-Gas can be operated safely using one line mode

Steady state and transient critical heat flux examinations

International Nuclear Information System (INIS)

Szabados, L.

1978-02-01

In steady state conditions within the P.W.R. parameter range the critical heat flux correlations based on local parameters reproduce the experimental data with less deviations than those based on system parameters. The transient experiments were restricted for the case of power transients. A data processing method for critical heat flux measurements has been developed and the applicability of quasi steady state calculation has been verified. (D.P.)

SCANAIR: A transient fuel performance code

International Nuclear Information System (INIS)

Moal, Alain; Georgenthum, Vincent; Marchand, Olivier

2014-01-01

Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

SCANAIR: A transient fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Moal, Alain, E-mail: alain.moal@irsn.fr; Georgenthum, Vincent; Marchand, Olivier

2014-12-15

Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

Development of a Two-dimensional Thermohydraulic Hot Pool Model and ITS Effects on Reactivity Feedback during a UTOP in Liquid Metal Reactors

International Nuclear Information System (INIS)

Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee

2009-01-01

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect

Calculations of transient fields in the Felix experiments at Argonne using null field integrated techniques

International Nuclear Information System (INIS)

Han, H.C.; Davey, K.R.; Turner, L.

1985-08-01

The transient eddy current problem is characteristically computationally intensive. The motivation for this research was to realize an efficient, accurate, solution technique involving small matrices via an eigenvalue approach. Such a technique is indeed realized and tested using the null field integral technique. Using smart (i.e., efficient, global) basis functions to represent unknowns in terms of a minimum number of unknowns, homogeneous eigenvectors and eigenvalues are first determined. The general excitatory response is then represented in terms of these eigenvalues/eigenvectors. Excellent results are obtained for the Argonne Felix cylinder experiments using a 4 x 4 matrix. Extension to the 3-D problem (short cylinder) is set up in terms of an 8 x 8 matrix

Experimental study and modelling of transient boiling

International Nuclear Information System (INIS)

Baudin, Nicolas

2015-01-01

A failure in the control system of the power of a nuclear reactor can lead to a Reactivity Initiated Accident in a nuclear power plant. Then, a power peak occurs in some fuel rods, high enough to lead to the coolant film boiling. It leads to an important increase of the temperature of the rod. The possible risk of the clad failure is a matter of interest for the Institut de Radioprotection et de Securite Nucleaire. The transient boiling heat transfer is not yet understood and modelled. An experimental set-up has been built at the Institut de Mecanique des Fluides de Toulouse (IMFT). Subcooled HFE-7000 flows vertically upward in a semi annulus test section. The inner half cylinder simulates the clad and is made of a stainless steel foil, heated by Joule effect. Its temperature is measured by an infrared camera, coupled with a high speed camera for the visualization of the flow topology. The whole boiling curve is studied in steady state and transient regimes: convection, onset of boiling, nucleate boiling, critical heat flux, film boiling and rewetting. The steady state heat transfers are well modelled by literature correlations. Models are suggested for the transient heat flux: the convection and nucleate boiling evolutions are self-similar during a power step. This observation allows to model more complex evolutions, as temperature ramps. The transient Hsu model well represents the onset of nucleate boiling. When the intensity of the power step increases, the film boiling begins at the same temperature but with an increasing heat flux. For power ramps, the critical heat flux decreases while the corresponding temperature increases with the heating rate. When the wall is heated, the film boiling heat transfer is higher than in steady state but it is not understood. A two-fluid model well simulates the cooling film boiling and the rewetting. (author)

Reactive transport modeling of the ABM experiment with Comsol Multiphysics

International Nuclear Information System (INIS)

Pekala, Marek; Idiart, Andres; Arcos, David

2012-01-01

Document available in extended abstract form only. The Swedish Organisation for Radioactive Waste Disposal (SKB) is considering disposal of the High Level Waste in a deep underground repository in a crystalline rock. According to the disposal concept, bentonite clay will be used in the near-field of the waste packages as buffer material. From solute transport point of view, the bentonite buffer is expected to provide a favourable environment, where radionuclide migration would be limited to slow diffusion and further retarded by sorption. In the KBS-3 repository design, the MX-80 bentonite is the reference buffer material. However, SKB has also been investigating alternative buffer materials. To this end, the field experiment Alternative Buffer Materials (ABM) was started at the Aespoe URL in 2006. Three packages of eleven different compacted bentonite blocks in different configurations have been tested over varying time scales. The packages with outer diameter of 0.28 m were deposited into 3 meter deep boreholes. After installation, packages were saturated and heated differently to target values. This contribution concerns the evolution of Package 1, which was initiated in December 2006 and ran for about 2.5 years. Post-mortem examination after retrieval showed that the initially contrasting chloride concentrations and cation-exchanger compositions between different bentonite blocks became significantly homogenised. It is thought that this behaviour could be explained as a first approximation by diffusion of major ions between the bentonite blocks coupled with cation-exchange. In this work, a modelling study to verify this hypothesis has been undertaken. In addition, the feasibility of implementing a reactive transport model into the Finite Element code COMSOL Multiphysics has been tested. The model considers a two-dimensional axisymmetric geometry of the depositional borehole, and includes coupled diffusion and cation-exchange of Na, K, Ca and Mg (as a chloride

Experimental study on transient boiling heat transfer

International Nuclear Information System (INIS)

Visentini, R.

2012-01-01

Boiling phenomena can be found in the everyday life, thus a lot of studies are devoted to them, especially in steady state conditions. Transient boiling is less known but still interesting as it is involved in the nuclear safety prevention. In this context, the present work was supported by the French Institute of Nuclear Safety (IRSN). In fact, the IRSN wanted to clarify what happens during a Reactivity-initiated Accident (RIA). This accident occurs when the bars that control the nuclear reactions break down and a high power peak is passed from the nuclear fuel bar to the surrounding fluid. The temperature of the nuclear fuel bar wall increases and the fluid vaporises instantaneously. Previous studies on a fuel bar or on a metal tube heated by Joule effect were done in the past in order to understand the rapid boiling phenomena during a RIA. However, the measurements were not really accurate because the measurement techniques were not able to follow rapid phenomena. The main goal of this work was to create an experimental facility able to simulate the RIA boiling conditions but at small scale in order to better understand the boiling characteristics when the heated-wall temperature increases rapidly. Moreover, the experimental set-up was meant to be able to produce less-rapid transients as well, in order to give information on transient boiling in general. The facility was built at the Fluid-Mechanics Institute of Toulouse. The core consists of a metal half-cylinder heated by Joule effect, placed in a half-annulus section. The inner half cylinder is made of a 50 microns thick stainless steel foil. Its diameter is 8 mm, and its length 200 mm. The outer part is a 34 mm internal diameter glass half cylinder. The semi-annular section is filled with a coolant, named HFE7000. The configuration allows to work in similarity conditions. The heated part can be place inside a loop in order to study the flow effect. The fluid temperature influence is taken into account as

Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

International Nuclear Information System (INIS)

Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

2004-01-01

As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

International Nuclear Information System (INIS)

Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

2000-03-01

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

Physical modelling of a rapid boron dilution transient

Energy Technology Data Exchange (ETDEWEB)

Anderson, N.G.; Hemstroem, B.; Karlsson, R. [Vattenfall Utveckling AB, Aelvkarleby (Sweden); Jacobson, S. [Vattenfall AB, Ringhals, Vaeroebacka (Sweden)

1995-09-01

The analysis of boron dilution accidents in pressurised water reactors has traditionally assumed that mixing is instantaneous and complete everywhere, eliminating in this way the possibility of concentration inhomogeneities. Situations can nevertheless arise where a volume of coolant with a low boron concentration may eventually enter the core and generate a severe reactivity transient. The work presented in this paper deals with a category of Rapid Boron Dilution Events characterised by a rapid start of a Reactor Coolant Pump (RCP) with a plug of relatively unborated water present in the RCS pipe. Model tests have been made at Vattenfall Utveckling AB in a simplified 1:5 scale model of a Westinghouse PWR. Conductivity measurements are used to determine dimensionless boron concentration. The main purpose of this experimental work is to define an experimental benchmark against which a mathematical model can be tested. The final goal is to be able to numerically predict Boron Dilution Transients. This work has been performed as a part of a Co-operative Agreement with Electricite` de France (EDF).

Response of steam-water mixtures to pressure transients

International Nuclear Information System (INIS)

Hull, L.M.

1985-01-01

During the transition phase of a hypothetical core-disruptive accident in a liquid-metal fast breeder reactor, melting fuel-steel mixtures may begin to boil, resulting in a two-phase mixture of molten reactor fuel and steel vapor. Dispersal of this mixture by pressure transients may prevent recriticality of the fuel material. This paper describes the results of a series of experiments that investigated the response of two-phase mixtures to pressure transients. Simulant fluids (steam/water) were used in a transparent 10.2-cm-dia, 63.5-cm-long acrylic tube. The pressure transient was provided by releasing pressurized nitrogen from a supply tank. The data obtained are in the form of pressure-time records and high-speed movies. The varied parameters are initial void fraction (10% and 40%) and transient pressure magnitude (3.45 and 310 kPa)

Current status of the transient integral fuel element performance code URANUS

International Nuclear Information System (INIS)

Preusser, T.; Lassmann, K.

1983-01-01

To investigate the behavior of fuel pins during normal and off-normal operation, the integral fuel rod code URANUS has been extended to include a transient version. The paper describes the current status of the program system including a presentation of newly developed models for hypothetical accident investigation. The main objective of current development work is to improve the modelling of fuel and clad material behavior during fast transients. URANUS allows detailed analysis of experiments until the onset of strong material transport phenomena. Transient fission gas analysis is carried out due to the coupling with a special version of the LANGZEIT-KURZZEIT-code (KfK). Fuel restructuring and grain growth kinetics models have been improved recently to better characterize pre-experimental steady-state operation; transient models are under development. Extensive verification of the new version has been carried out by comparison with analytical solutions, experimental evidence, and code-to-code evaluation studies. URANUS, with all these improvements, has been successfully applied to difficult fast breeder fuel rod analysis including TOP, LOF, TUCOP, local coolant blockage and specific carbide fuel experiments. Objective of further studies is the description of transient PCMI. It is expected that the results of these developments will contribute significantly to the understanding of fuel element structural behavior during severe transients. (orig.)

Modeling of oxygen gas diffusion and consumption during the oxic transient in a disposal cell of radioactive waste

International Nuclear Information System (INIS)

De Windt, Laurent; Marsal, François; Corvisier, Jérôme; Pellegrini, Delphine

2014-01-01

Highlights: • This paper deals with the geochemistry of underground HLW disposals. • The oxic transient is a key issue in performance assessment (e.g. corrosion, redox). • A reactive transport model is explicitly coupled to gas diffusion and reactivity. • Application to in situ experiment (Tournemire laboratory) and HLW disposal cell. • Extent of the oxidizing/reducing front is investigated by sensitivity analysis. - Abstract: The oxic transient in geological radioactive waste disposals is a key issue for the performance of metallic components that may undergo high corrosion rates under such conditions. A previous study carried out in situ in the argillite formation of Tournemire (France) has suggested that oxic conditions could have lasted several years. In this study, a multiphase reactive transport model is performed with the code HYTEC to analyze the balance between the kinetics of pyrite oxidative dissolution, the kinetics of carbon steel corrosion and oxygen gas diffusion when carbon steel components are emplaced in the geological medium. Two cases were modeled: firstly, the observations made in situ have been reproduced, and the model established was then applied to a disposal cell for high-level waste (HLW) in an argillaceous formation, taking into account carbon steel components and excavated damaged zones (EDZ). In a closed system, modeling leads to a complete and fast consumption of oxygen in both cases. Modeling results are more consistent with the in situ test while considering residual voids between materials and/or a water unsaturated state allowing for oxygen gas diffusion (open conditions). Under similar open conditions and considering ventilation of the handling drifts, a redox contrast occurs between reducing conditions at the back of the disposal cell (with anoxic corrosion of steel and H 2 production) and oxidizing conditions at the front of the cell (with oxic corrosion of steel). The extent of the oxidizing/reducing front in the

Analysis of a main steam isolation value closure anticipated transient without scram in a boiling water reactor

International Nuclear Information System (INIS)

Liaw, T.J.; Pan, C.; Chen, G.S.

1989-01-01

Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main steam isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4. Without boron injection and makeup coolant, the reactor loses its coolant inventory very quickly and the reactor power drops rapidly to ∼ 16% of rated power due to negative void reactivity. With coolant makeup from the high-pressure core spray and the reactor core isolation cooling systems, the rector reaches a quasi-steady-state condition after an initially rapidly changing transient. The dome pressure, downcomer water level, and core power oscillate around a mean value; the average core power is ∼ 15%, which is approximately equal to the power needed to heat and evaporate the subcooled makeup coolant. Lower boron concentrations in the core tend to complicate reactor behavior due to the combination of two competing phenomena: the negative boron reactivity and the positive reactivity caused by a void collapse

RAP-2A Computer code for transients analysis in fast reactors

International Nuclear Information System (INIS)

Iftode, I.; Popescu, C.; Turcu, I.; Biro, L.

1975-10-01

The RAP-2A computer code is designed for analyzing thermohydraulic transients and/or steady state problems for large LMFBR cores. Physical and mathematical models, main input-output data, the flow chart of the code and a sample problem are given. RAP-2A calculates the power and the thermoydraulic transients initiated by a flow or reactivity changes, from a normal operating state of the reactor up to core disassembly. In this analysis a representative fuel pin is considered: a one-group space-independent (point) kinetics model to describe the neutron kinetics and a one-dimensional model describing the heat transfer (radial in the fuel and axial in the coolant) are used. Mechanical deformations due to temperature gradient, pressure losses, fuel melting, etc., are also calculated. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

Keeping the secret: Insights from repeated catchment-scale tracer experiments under transient conditions

Science.gov (United States)

Bogner, Christina; Hauhs, Michael; Lange, Holger

2016-04-01

Catchment-level tracer experiments are generally performed to identify site-specific hydrological response functions of the catchment. The existence and uniqueness of these response functions are hardly ever questioned. Here, we report on a series of replicated tracer experiments in two small first-order catchments, G1 (0.6 ha, roofed) and F4 (2.3 ha, without roof) at Gårdsjön in SW Sweden. The soils in both catchments are shallow (500 m2) the experiments were done without a roof mostly at transient conditions. The catchment F4 was equipped with a sprinkler system with a watering capacity of around 38-45 m3 day-1. Natural rainfall comes in addition. A bromide tracer solution was injected to groundwater at a single location about 40 m upstream the weir over a period of less than an hour, and was monitored using a set of groundwater tubes and the weir at the outlet over the following 4 days. In addition, discharge was measured. The experiments were repeated each summer from 2007 to 2015. While steady state conditions were guaranteed in G1, steady runoff has been achieved only four times in F4. We investigated tracer recovery rates against cumulated runoff since tracer application. Substantially different transit times and qualitatively different behaviour of the breakthrough curves were observed, even under steady state conditions. In G1, no single system response function could be identified in 5 replicates. Similarly, the catchment response functions in F4 under steady state differed between experiments. However, they remained in a similar range as in G1. Based on these results, we question the identifiability of flow paths and system properties, such as saturated water content or hydrologic transmissivity, at the catchment scale using tracer experiments. Rather, the series demonstrate the utter importance of the initial and boundary conditions which largely determine the response of the system to inert tracer pulses.

The use of single-crystal iron frames in transient field measurements

International Nuclear Information System (INIS)

Zalm, P.C.; Laan, J. van der; Middelkoop, G. van

1979-01-01

Single-crystal Fe frames have been investigated for use as a ferromagnetic backing in transient magnetic field experiments. For this purpose the surface magnetization as a function of applied magnetic field has been determined with the magneto-optical Kerr effect. The frames, which have two sides parallel to the crystal axis, can be fully magnetized at low external fields such that fringing fields are negligibly small. These single-crystal Fe backings have been used in several transient magnetic field experiments. Comparison of the measured precession angles with previous results, obtained in polycrystalline Fe foils at high external magnetic fields, shows that the single-crystal backings are satisfactory. After extended periods of heavy-ion bombardment the crystals exhibited no radiation damage effects. The absence of fringing fields leads to a reduction of a factor of four in the measuring time for transient field experiments. (Auth.)

Investigations of anticipated transients without scram (ATWS) for the high temperature reactor

International Nuclear Information System (INIS)

Heckhoff, H.D.

1981-10-01

In this study anticipated transients without scram (ATWS) are investigated for the high temperature reactor, especially for the thorium high temperature reactor (THTR) 300 MWe as an example. It is shown that the two ATWS 'feedwater flow reduction from full power' and 'positive reactivity insertion of 1 mNile/s from 40 per cent power' are the most important transients for the THTR. The additional load caused by the ATWS can be reduced sufficiently by some small modifications of the afterheat removal system. Supplementary precautions are not necessary. In the last part of this study some possibilities to improve the behaviour of the power plant are shown with regard to high temperature reactors of the future, the partial scram as well as some modifications of heating and cooling of the steam generator. (orig.) [de

Energy performance of a micro-cogeneration device during transient and steady-state operation: Experiments and simulations

International Nuclear Information System (INIS)

Rosato, Antonio; Sibilio, Sergio

2013-01-01

Micro-cogeneration is a well-established technology and its deployment has been considered by the European Community as one of the most effective measure to save primary energy and to reduce greenhouse gas emissions. As a consequence, the estimation of the potential impact of micro-cogeneration devices is necessary to design policy and to energetically, ecologically and economically rank these systems among other potential energy saving and CO 2 -reducing measures. Even if transient behaviour can be very important when the engine is frequently started and stopped and allowed to cool-down in between, for the sake of simplicity mainly static and simplified methods are used for assessing the performance of cogeneration devices, completely neglecting the dynamic response of the units themselves. In the first part of this paper a series of experiments is illustrated and discussed in detail in order to highlight and compare the transient and stationary operation of a natural gas fuelled reciprocating internal combustion engine based cogeneration unit with 6.0 kW as nominal electric output and 11.7 kW as nominal thermal output. The measured performance of the cogeneration device is also compared with the performance of the system calculated on the basis of the efficiency values suggested by the manufacturer in order to highlight and quantify the discrepancy between the two approaches in evaluating the unit operation. Finally the experimental data are also compared with those predicted by a simulation model developed within IEA/ECBCS Annex 42 and experimentally calibrated by the authors in order to assess the model reliability for studying and predicting the performance of the system under different operating scenarios. -- Highlights: ► Transient operation of a cogeneration system has been experimentally investigated. ► Steady-state operation of a cogeneration device has been experimentally evaluated. ► Measured data have been compared with those predicted by a

PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments

International Nuclear Information System (INIS)

Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella

2005-01-01

Full text of publication follows: In the field of PWR reactor safety, one of the topics that is currently of major interest worldwide is that of inadvertent boron dilution events. The safety issue involved in such scenarios is that inadvertent transport into the reactor core of un-borated water - or water having only a low boron concentration - can lead to local recriticality and possibly to power excursions. Studies on various accidental sequences that could initiate boron dilution events revealed that some SBLOCAs, occurring in the primary system, lead to reflux condenser conditions and subsequent re-establishment of natural circulation are of particular significance. In this work, the first field of analysis is related to the investigation of the thermal - hydraulic conditions that could lead to boron dilution events, such as the stop of natural circulation within primary system and the subsequent start of reflux condenser functioning mode. The investigation of the primary thermal - hydraulic conditions has been performed using the experimental results obtained in the PKL test integral facility in which some SBLOCA sequences have been carried out. Particular useful were the PKL III E experiments data whose results have been numerically reproduced using the code Relap5/MOD3.3/Beta code, contributing to understand the complex thermalhydraulic phenomena related to a PWR boron dilution event. The second field of analysis is related to the effects that possible displacements of un-borated water slugs towards the Reactor Pressure Vessel (RPV) could have on the core reactivity. A numerical approach using the Relap5 reactor kinetics model has been adopted to integrate the experimental thermal - hydraulic data obtained in the PKL III E tests. A careful analysis has been performed in order to establish which core conditions at incident start could produce the largest reactivity increase as a consequence of restarting of natural circulation during the primary system

PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments

Energy Technology Data Exchange (ETDEWEB)

Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella [Dipartimento di Ingegneria Nucleare, Viale delle Scienze, 90128 Palermo (Italy)

2005-07-01

Full text of publication follows: In the field of PWR reactor safety, one of the topics that is currently of major interest worldwide is that of inadvertent boron dilution events. The safety issue involved in such scenarios is that inadvertent transport into the reactor core of un-borated water - or water having only a low boron concentration - can lead to local recriticality and possibly to power excursions. Studies on various accidental sequences that could initiate boron dilution events revealed that some SBLOCAs, occurring in the primary system, lead to reflux condenser conditions and subsequent re-establishment of natural circulation are of particular significance. In this work, the first field of analysis is related to the investigation of the thermal - hydraulic conditions that could lead to boron dilution events, such as the stop of natural circulation within primary system and the subsequent start of reflux condenser functioning mode. The investigation of the primary thermal - hydraulic conditions has been performed using the experimental results obtained in the PKL test integral facility in which some SBLOCA sequences have been carried out. Particular useful were the PKL III E experiments data whose results have been numerically reproduced using the code Relap5/MOD3.3/Beta code, contributing to understand the complex thermalhydraulic phenomena related to a PWR boron dilution event. The second field of analysis is related to the effects that possible displacements of un-borated water slugs towards the Reactor Pressure Vessel (RPV) could have on the core reactivity. A numerical approach using the Relap5 reactor kinetics model has been adopted to integrate the experimental thermal - hydraulic data obtained in the PKL III E tests. A careful analysis has been performed in order to establish which core conditions at incident start could produce the largest reactivity increase as a consequence of restarting of natural circulation during the primary system

Development and verification of an efficient spatial neutron kinetics method for reactivity-initiated event analyses

International Nuclear Information System (INIS)

Ikeda, Hideaki; Takeda, Toshikazu

2001-01-01

A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasistatic (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data. (author)

Reactivation of Rate Remapping in CA3.

Science.gov (United States)

Schwindel, C Daniela; Navratilova, Zaneta; Ali, Karim; Tatsuno, Masami; McNaughton, Bruce L

2016-09-07

The hippocampus is thought to contribute to episodic memory by creating, storing, and reactivating patterns that are unique to each experience, including different experiences that happen at the same location. Hippocampus can combine spatial and contextual/episodic information using a dual coding scheme known as "global" and "rate" remapping. Global remapping selects which set of neurons can activate at a given location. Rate remapping readjusts the firing rates of this set depending on current experience, thus expressing experience-unique patterns at each location. But can the experience-unique component be retrieved spontaneously? Whereas reactivation of recent, spatially selective patterns in hippocampus is well established, it is never perfect, raising the issue of whether the experiential component might be absent. This question is key to the hypothesis that hippocampus can assist memory consolidation by reactivating and broadcasting experience-specific "index codes" to neocortex. In CA3, global remapping exhibits attractor-like dynamics, whereas rate remapping apparently does not, leading to the hypothesis that only the former can be retrieved associatively and casting doubt on the general consolidation hypothesis. Therefore, we studied whether the rate component is reactivated spontaneously during sleep. We conducted neural ensemble recordings from CA3 while rats ran on a circular track in different directions (in different sessions) and while they slept. It was shown previously that the two directions of running result in strong rate remapping. During sleep, the most recent rate distribution was reactivated preferentially. Therefore, CA3 can retrieve patterns spontaneously that are unique to both the location and the content of recent experience. The hippocampus is required for memory of events and their spatial contexts. The primary correlate of hippocampal activity is location in space, but multiple memories can occur in the same location. To be useful

Use of reactivity constraints for the automatic control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.; Lanning, D.D.; Ray, A.

1985-01-01

A theoretical framework for the automatic control of reactor power has been developed and experimentally evaluated on the 5 MWt Research Reactor that is operated by the Massachusetts Institute of Technology. The controller functions by restricting the net reactivity so that it is always possible to make the reactor period infinite at the desired termination point of a transient by reversing the direction of motion of whatever control mechanism is associated with the controller. This capability is formally designated as ''feasibility of control''. It has been shown experimentally that maintenance of feasibility of control is a sufficient condition for the automatic control of reactor power. This research should be of value in the design of closed-loop controllers, in the creation of reactivity displays, in the provision of guidance to operators regarding the timing of reactivity changes, and as an experimental envelope within which alternate control strategies can be evaluated

MTR fuel element burn-up measurements by the reactivity method

International Nuclear Information System (INIS)

Zuniga, A.; Cuya, T.R.; Ravnik, M.

2003-01-01

Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

A randomized controlled trial to test the effect of multispecies probiotics on cognitive reactivity to sad mood

NARCIS (Netherlands)

Steenbergen, L.; Sellaro, R.; van Hemert, S.; Bosch, J.A.; Colzato, L.S.

2015-01-01

Background: Recent insights into the role of the human microbiota in cognitive and affective functioning have led to the hypothesis that probiotic supplementation may act as an adjuvant strategy to ameliorate or prevent depression. Objective: Heightened cognitive reactivity to normal, transient

Detailed characterizations of a Comparative Reactivity Method (CRM) instrument: experiments vs. modelling

Science.gov (United States)

Michoud, V.; Hansen, R. F.; Locoge, N.; Stevens, P. S.; Dusanter, S.

2015-04-01

simple chemical mechanism, taking into account the inorganic chemistry from IUPAC 2001 and a simple organic chemistry scheme including only a generic RO2 compounds for all oxidized organic trace gases; and (2) a more exhaustive chemical mechanism, based on the Master Chemical Mechanism (MCM), including the chemistry of the different trace gases used during laboratory experiments. Both mechanisms take into account self- and cross-reactions of radical species. The simulations using these mechanisms allow reproducing the magnitude of the corrections needed to account for NO interferences and a deviation from pseudo first-order kinetics, as well as their dependence on the Pyrrole-to-OH ratio and on bimolecular reaction rate constants of trace gases. The reasonable agreement found between laboratory experiments and model simulations gives confidence in the parameterizations proposed to correct the Total OH reactivity measured by CRM. However, it must be noted that the parameterizations presented in this paper are suitable for the CRM instrument used during the laboratory characterization and may be not appropriate for other CRM instruments, even if similar behaviours should be observed. It is therefore recommended that each group characterizes its own instrument following the recommendations given in this study. Finally, the assessment of the limit of detection and total uncertainties is discussed and an example of field deployment of this CRM instrument is presented.

Modeling and analysis of thermal-hydraulic response of uranium-aluminum reactor fuel plates under transient heatup conditions

Energy Technology Data Exchange (ETDEWEB)

Navarro-Valenti, S.; Kim, S.H.; Georgevich, V. [Oak Ridge National Lab., TN (United States)] [and others

1995-09-01

The purpose of this paper is to describe the analysis performed to predict the thermal behavior of fuel miniplates under rapid transient heatup conditions. The possibility of explosive boiling was considered, and it was concluded that the heating rates are not large enough for explosive boiling to occur. However, transient boiling effects were pronounced. Because of the complexity of transient pool boiling and the unavailability of experimental data for the situations studied, an approximation was made that predicted the data very well within the uncertainties present. If pool boiling from the miniplates had been assumed to be steady during the heating pulse, the experimental data would have been greatly overestimated. This fact demonstrates the importance of considering the transient nature of heat transfer in the analysis of reactivity excursion accidents. An additional contribution of the present work is that it provided data on highly subcooled steady nulceate boiling from the cooling portion of the thermocouple traces.

Critical experiment program of heterogeneous core composed for LWR fuel rods and low enriched uranyl nitrate solution

International Nuclear Information System (INIS)

Miyoshi, Yoshinori; Yamamoto, Toshihiro; Watanabe, Shouichi; Nakamura, Takemi

2003-01-01

In order to stimulate the criticality characteristics of a dissolver in a reprocessing plant, a critical experiment program of heterogeneous cores is under going at a Static Critical Experimental Facility, STACY in Japan Atomic Energy Research Institute, JAERI. The experimental system is composed of 5w/o enriched PWR-type fuel rod array immersed in 6w/o enriched uranyl nitrate solution. First series of experiments are basic benchmark experiments on fundamental critical data in order to validate criticality calculation codes for 'general-form system' classified in the Japanese Criticality Safety Handbook, JCSHB. Second series of experiments are concerning the neutron absorber effects of fission products related to the burn-up credit Level-2. For demonstrating the reactivity effects of fission products, reactivity effects of natural elements such as Sm, Nd, Eu and 103 Rh, 133 Cs, solved in the nitrate solution are to be measured. The objective of third series of experiments is to validate the effect of gadolinium as a soluble neutron poison. Properties of temperature coefficients and kinetic parameters are also studied, since these parameters are important to evaluate the transient behavior of the criticality accident. (author)

PWR [pressurized water reactor] pressurizer transient response: Final report

International Nuclear Information System (INIS)

Murphy, S.I.

1987-08-01

To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

A review of experiments and results from the TREAT facility

International Nuclear Information System (INIS)

Deitrich, L.W.; Dickerman, C.E.; Klickman, A.E.; Wright, A.E.

1998-01-01

The Transient Reactor Test (TREAT) facility was designed and built in the late 1950s at Argonne National Laboratory (ANL) to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off-normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light water reactor (LWR) elements in a steam environment to obtain fission product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

International Nuclear Information System (INIS)

Peng Hong Liem; Surian Pinem; Tagor Malem Sembiring; Tran Hoai Nam

2015-01-01

A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

Ghrelin potentiates cardiac reactivity to stress by modulating sympathetic control and beta-adrenergic response.

Science.gov (United States)

Camargo-Silva, Gabriel; Turones, Larissa Córdova; da Cruz, Kellen Rosa; Gomes, Karina Pereira; Mendonça, Michelle Mendanha; Nunes, Allancer; de Jesus, Itamar Guedes; Colugnati, Diego Basile; Pansani, Aline Priscila; Pobbe, Roger Luis Henschel; Santos, Robson; Fontes, Marco Antônio Peliky; Guatimosim, Silvia; de Castro, Carlos Henrique; Ianzer, Danielle; Ferreira, Reginaldo Nassar; Xavier, Carlos Henrique

2018-03-01

Prior evidence indicates that ghrelin is involved in the integration of cardiovascular functions and behavioral responses. Ghrelin actions are mediated by the growth hormone secretagogue receptor subtype 1a (GHS-R1a), which is expressed in peripheral tissues and central areas involved in the control of cardiovascular responses to stress. In the present study, we assessed the role of ghrelin - GHS-R1a axis in the cardiovascular reactivity to acute emotional stress in rats. Ghrelin potentiated the tachycardia evoked by restraint and air jet stresses, which was reverted by GHS-R1a blockade. Evaluation of the autonomic balance revealed that the sympathetic branch modulates the ghrelin-evoked positive chronotropy. In isolated hearts, the perfusion with ghrelin potentiated the contractile responses caused by stimulation of the beta-adrenergic receptor, without altering the amplitude of the responses evoked by acetylcholine. Experiments in isolated cardiomyocytes revealed that ghrelin amplified the increases in calcium transient changes evoked by isoproterenol. Taken together, our results indicate that the Ghrelin-GHS-R1a axis potentiates the magnitude of stress-evoked tachycardia by modulating the autonomic nervous system and peripheral mechanisms, strongly relying on the activation of cardiac calcium transient and beta-adrenergic receptors. Copyright © 2018 Elsevier Inc. All rights reserved.

OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Third Workshop (V1000-CT3)

International Nuclear Information System (INIS)

2005-01-01

The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. The technical topics presented at this workshop were: Review of the benchmark activities after the 2. Workshop; - Discussion of participant's feedback and introduced modifications

Positive void reactivity

International Nuclear Information System (INIS)

Diamond, D.J.

1992-09-01

This report is a review of some of the important aspects of the analysis of large loss-of-coolant accidents (LOCAs). One important aspect is the calculation of positive void reactivity. To study this subject the lattice physics codes used for void worth calculations and the coupled neutronic and thermal-hydraulic codes used for the transient analysis are reviewed. Also reviewed are the measurements used to help validate the codes. The application of these codes to large LOCAs is studied with attention focused on the uncertainty factor for the void worth used to bias the results. Another aspect of the subject dealt with in the report is the acceptance criteria that are applied. This includes the criterion for peak fuel enthalpy and the question of whether prompt criticality should also be a criterion. To study the former, fuel behavior measurements and calculations are reviewed. (Author) (49 refs., 2 figs., tab.)

Experimental investigations on the transient behaviour of nuclear heat plants with natural convection

International Nuclear Information System (INIS)

Adam, E.; Sydow, J.; Wolff, J.

1988-01-01

Apart from the theoretical approach, practical experiments concerning the transient behaviour of the primary loop of reactors with natural coolant convection are necessary in order to evaluate the safety systems of reactors providing heat for industrial and communal consumers. The article presents experiments concerning the transient behaviour of the experimental plant DANTON, which models the reactor AST-500, and gives a preview of further research. (orig.) [de

Instructions for applying inverse method for reactivity measurement

International Nuclear Information System (INIS)

Milosevic, M.

1988-11-01

This report is a brief description of the completed method for reactivity measurement. It contains description of the experimental procedure needed instrumentation and computer code IM for determining reactivity. The objective of this instructions manual is to enable experiments and reactivity measurement on any critical system according to the methods adopted at the RB reactor

Transients in reactors for power systems compensation

Science.gov (United States)

Abdul Hamid, Haziah

This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

Narcisismo reativo e experiência religiosa contemporânea: culpa substituída pela vergonha? Reactive narcissism and contemporary religious experience: a shift from guilt to shame?

Directory of Open Access Journals (Sweden)

Mary Rute Gomes Esperandio

2007-08-01

Full Text Available O texto apresenta um recorte da reflexão desenvolvida na tese de doutorado: "Narcisismo e sacrifício: Modo de subjetivação e religiosidade contemporânea" (Esperandio, 2006. Trata-se de uma análise da experiência religiosa promovida pela Igreja Universal do Reino de Deus (IURD a partir da sua proposta de sacrifício. A abordagem trabalha com a concepção de narcisismo ativo e reativo e defende a idéia de que a prática do sacrifício, tal como proposta pela IURD, articula elementos do narcisismo reativo e estabelece-se como uma tecnologia do eu, usada como remédio para lidar com a experiência contemporânea da vergonha.This text presents part of the reflection developed in my Doctoral Thesis on "Narcissism and sacrifice: Mode of subjectivation and contemporary religiosity" (Esperandio, 2006. It analyses the religious experience supported by the Universal Church of the Kingdom of God (UCKG on its proposal of sacrifice. This approach sustains the conception of active and reactive narcissism and defends the idea that such practice of sacrifice proposed by UCKG relates some elements of reactive narcissism and establishes itself as a Technology of Self used as a remedy to deal with the contemporary experience of shame.

Performance of neutron kinetics models for ADS transient analyses

International Nuclear Information System (INIS)

Rineiski, A.; Maschek, W.; Rimpault, G.

2002-01-01

can also apply this approach for estimating errors of point-kinetics simulations or for ameliorating the employed point-kinetics models. Though the performance of the point-kinetics model can be insufficient in the subcritical case, the quasi-static approach is still valid if the shape steps are chosen properly. It is worthwhile to mention that in combination with properly computed correction factor tables, one can use the reactivity and power distributions obtained for 'critical' reactor models; this approach can simplify ADS-related application of conventional accident analyses codes (developed in the past for transient analyses of critical reactors). However, for analyzing severe transients in ADSs, which involve gross core material configuration changes, one can hardly avoid using of space-time kinetics methods, this holds similarly for critical reactor systems. (authors)

Transcriptome wide annotation of eukaryotic RNase III reactivity and degradation signals.

Directory of Open Access Journals (Sweden)

Jules Gagnon

2015-02-01

Full Text Available Detection and validation of the RNA degradation signals controlling transcriptome stability are essential steps for understanding how cells regulate gene expression. Here we present complete genomic and biochemical annotations of the signals required for RNA degradation by the dsRNA specific ribonuclease III (Rnt1p and examine its impact on transcriptome expression. Rnt1p cleavage signals are randomly distributed in the yeast genome, and encompass a wide variety of sequences, indicating that transcriptome stability is not determined by the recurrence of a fixed cleavage motif. Instead, RNA reactivity is defined by the sequence and structural context in which the cleavage sites are located. Reactive signals are often associated with transiently expressed genes, and their impact on RNA expression is linked to growth conditions. Together, the data suggest that Rnt1p reactivity is triggered by malleable RNA degradation signals that permit dynamic response to changes in growth conditions.

Transcriptome Wide Annotation of Eukaryotic RNase III Reactivity and Degradation Signals

Science.gov (United States)

Gagnon, Jules; Lavoie, Mathieu; Catala, Mathieu; Malenfant, Francis; Elela, Sherif Abou

2015-01-01

Detection and validation of the RNA degradation signals controlling transcriptome stability are essential steps for understanding how cells regulate gene expression. Here we present complete genomic and biochemical annotations of the signals required for RNA degradation by the dsRNA specific ribonuclease III (Rnt1p) and examine its impact on transcriptome expression. Rnt1p cleavage signals are randomly distributed in the yeast genome, and encompass a wide variety of sequences, indicating that transcriptome stability is not determined by the recurrence of a fixed cleavage motif. Instead, RNA reactivity is defined by the sequence and structural context in which the cleavage sites are located. Reactive signals are often associated with transiently expressed genes, and their impact on RNA expression is linked to growth conditions. Together, the data suggest that Rnt1p reactivity is triggered by malleable RNA degradation signals that permit dynamic response to changes in growth conditions. PMID:25680180

Study of the linearity of CABRI experimental ionization chambers during RIA transients

Science.gov (United States)

Lecerf, J.; Garnier, Y.; Hudelot, JP.; Duc, B.; Pantera, L.

2018-01-01

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center and funded by the French Nuclear Safety and Radioprotection Institute (IRSN). For the purpose of the CABRI International Program (CIP), operated and managed by IRSN under an OECD/NEA framework it has been refurbished since 2003 to be able to provide experiments in prototypical PWR conditions (155 bar, 300 °C) in order to study the fuel behavior under Reactivity Initiated Accident (RIA) conditions. This paper first reminds the objectives of the power commissioning tests performed on the CABRI facility. The design and location of the neutron detectors monitoring the core power are also presented. Then it focuses on the different methodologies used to calibrate the detectors and check the consistency and co-linearity of the measurements. Finally, it presents the methods used to check the linearity of the neutron detectors up to the high power levels ( 20 GW) reached during power transients. Some results obtained during the power tests campaign are also presented.

Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method

Energy Technology Data Exchange (ETDEWEB)

He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)

2015-12-15

Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.

M. I. T. studies of transient X-ray phenomena

Energy Technology Data Exchange (ETDEWEB)

Canizares, C R [Massachusetts Inst. of Tech., Cambridge (USA). Dept. of Physics

1976-06-01

A variety of transient X-ray phenomena have been studied by the M.I.T. X-ray Astronomy Group. Data from the OSO-7 satellite reveal both long and short time-scale transients. Extensive observations have been made of the Lupus X-ray Nova (3U1543-47) and of GX339-4 (MX1658-48) which may represent a very different type of transient source. A unique, intense X-ray flare lasting ten minutes was also recorded, and the X-ray emission from the active galaxy Cen A was found to vary significantly over a period of several days. In a recent balloon flight the Crab pulsar, NP0532, was observed to exhibit a transient pulsed component distinct from the usual main pulse and interpulse. A sounding-rocket experiment detected an ultrasoft transient X-ray source tentatively associated with SS Cygni, and preliminary results from SAS-3 show a very hard spectrum for the new source A0535+26. On the other hand, extensive OSO-7 null observations of both Type I and II supernovae and of the flaring radio star Algol make it unlikely that these types of objects are potent transient X-ray emitters.

Operation databook of the fuel treatment system of the Static Experiment Critical Facility (STACY) and the Transient Experiment Critical Facility (TRACY). JFY 2004 to JFY 2008

International Nuclear Information System (INIS)

Kokusen, Junya; Sumiya, Masato; Seki, Masakazu; Kobayashi, Fuyumi; Ishii, Junichi; Umeda, Miki

2013-02-01

Uranyl nitrate solution fuel used in the Static Experiment Critical Facility (STACY) and the Transient Experiment Critical Facility (TRACY) is adjusted in the Fuel Treatment System, in which such parameters are varied as concentration of uranium, free nitric acid, soluble neutron poison, and so on. Operations for concentration and denitration of the solution fuel were carried out with an evaporator from JFY 2004 to JFY 2008 in order to adjust the fuel to the experimental condition of the STACY and the TRACY. In parallel, the solution fuel in which some kinds of soluble neutron poison were doped was also adjusted in JFY 2005 and JFY 2006 for the purpose of the STACY experiments to determine neutron absorption effects brought by fission products, etc. After these experiments in the STACY, a part of the solution fuel including the soluble neutron poison was purified by the solvent extraction method with mixer-settlers in JFY 2006 and JFY 2007. This report summarizes operation data of the Fuel Treatment System from JFY 2004 to JFY 2008. (author)

Recurrent and Transient Spinal Pain Among Commercial Helicopter Pilots.

Science.gov (United States)

Andersen, Knut; Baardsen, Roald; Dalen, Ingvild; Larsen, Jan Petter

2015-11-01

The aim of this study was to provide information on the occurrence of spinal pain, i.e., low back and neck pain, among commercial helicopter pilots, along with possible associations between pain and anthropometric and demographic factors and flying exposure. Data were collected through a subjective and retrospective survey among all the 313 (294 men, 19 women) full-time pilots employed by two helicopter companies. A questionnaire was used to assess the extent of spinal complaints in a transient and recurrent pain pattern along with information on physical activities, occupational flying experience, and airframes. The survey had 207 responders (194 men, 13 women). The pilots had extensive flying experience. Spinal pain was reported by 67%. Flying-related transient pain was reported among 50%, whereas recurrent spinal pain, not necessarily associated with flying, was reported by 52%. Women experienced more pain, but sample size prevented further conclusions. Male pilots reporting any spinal pain flew significantly more hours last year (median 500 h, IQR 400-650) versus men with no pain (median 445 h, IQR 300-550). Male pilots with transient or recurrent spinal pain did not differ from nonaffected male colleagues in the measured parameters. Spinal pain is a frequent problem among male and female commercial helicopter pilots. For men, no significant associations were revealed for transient or recurrent spinal pain with age, flying experience in years, total hours, annual flying time, type of aircraft, or anthropometric factors except for any spinal pain related to hours flown in the last year.

Application of a combined superconducting fault current limiter and STATCOM to enhancement of power system transient stability

Energy Technology Data Exchange (ETDEWEB)

Mahdad, Belkacem, E-mail: bemahdad@mselab.org; Srairi, K.

2013-12-15

Highlights: •A simple interactive model SFCL–STATCOM Controller is proposed to enhance the transient stability. •The STATCOM controller is integrated in coordination with the SFCL to support the excessive reactive power during fault. •Voltage stability index based continuation power flow is used to locate the STATCOM and the SFCL. •The clearing time improved compared to other cases (with only SFCL, with only STATCOM). •The choice of the STATCOM parameters is very important to exploit efficiently the integration of STATCOM Controller. -- Abstract: Stable and reliable operation of the power system network is dependent on the dynamic equilibrium between energy production and power demand under large disturbance such as short circuit or important line tripping. This paper investigates the use of combined model based superconducting fault current limiter (SFCL) and shunt FACTS Controller (STATCOM) for assessing the transient stability of a power system considering the automatic voltage regulator. The combined model located at a specified branch based on voltage stability index using continuation power flow. The main role of the proposed combined model is to achieve simultaneously a flexible control of reactive power using STATCOM Controller and to reduce fault current using superconducting technology based SFCL. The proposed combined model has been successfully adapted within the transient stability program and applied to enhance the transient power system stability of the WSCC9-Bus system. Critical clearing time (CCT) has been used as an index to evaluate and validate the contribution of the proposed coordinated Controller. Simulation results confirm the effectiveness and perspective of this combined Controller to enhance the dynamic power system performances.

Oxide fuel pin transient performance analysis and design with the TEMECH code

International Nuclear Information System (INIS)

Bard, F.E.; Dutt, S.P.; Hinman, C.A.; Hunter, C.W.; Pitner, A.L.

1986-01-01

The TEMECH code is a fast-running, thermal-mechanical-hydraulic, analytical program used to evaluate the transient performance of LMR oxide fuel pins. The code calculates pin deformation and failure probability due to fuel-cladding differential thermal expansion, expansion of fuel upon melting, and fission gas pressurization. The mechanistic fuel model in the code accounts for fuel cracking, crack closure, porosity decrease, and the temperature dependence of fuel creep through the course of the transient. Modeling emphasis has been placed on results obtained from Fuel Cladding Transient Test (FCTT) testing, Transient Fuel Deformation (TFD) tests and TREAT integral fuel pin experiments

LOFT facility PSS experiments: analysis of wet well vertical loads resulting from transient initiation

International Nuclear Information System (INIS)

Berta, V.T.

1977-05-01

Fourteen experiments on the Loss-of-Fluid Test (LOFT) facility pressure suppression system (PSS) are analyzed in relation to the vertical load generated on the suppression tank in the first 0.5 sec of the transient. Variations in principle parameters affecting the generation of vertical loads were included in the experiments. The internal and external vent submergences are identified from the analysis as being parameters which are first order in influencing the magnitude of the vertical load. These parameters are geometric in nature and depend only on PSS design. Physical parameters of total energy input and rate of energy input to the dry well, which influence the dry well pressurization, also are identified as being first order in influencing the magnitude of the vertical loads. The vertical load magnitude is a direct function of these geometric and physical parameters. The analysis indicates that a small value in any one of the parameters will cause the vertical load to be small and to have little dependence on the magnitude of the other parameters. In addition, the phenomena of nonuniform nonsynchronized vent inlet pressures, which have origins that are either geometric, physical, or a combination of both, act as a significant vertical load reduction mechanism

Thermal-Hydraulic Analyses of Transients in an Actinide-Burner Reactor Cooled by Forced Convection of Lead Bismuth

Energy Technology Data Exchange (ETDEWEB)

Davis, Cliff Bybee

2003-09-01

The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.

Platelet activation, function, and reactivity in atherosclerotic carotid artery stenosis: a systematic review of the literature.

LENUS (Irish Health Repository)

Kinsella, J A

2012-09-27

An important proportion of transient ischemic attack or ischemic stroke is attributable to moderate or severe (50-99%) atherosclerotic carotid stenosis or occlusion. Platelet biomarkers have the potential to improve our understanding of the pathogenesis of vascular events in this patient population. A detailed systematic review was performed to collate all available data on ex vivo platelet activation and platelet function\\/reactivity in patients with carotid stenosis. Two hundred thirteen potentially relevant articles were initially identified; 26 manuscripts met criteria for inclusion in this systematic review. There was no consistent evidence of clinically informative data from urinary or soluble blood markers of platelet activation in patients with symptomatic moderate or severe carotid stenosis who might be considered suitable for carotid intervention. Data from flow cytometry studies revealed evidence of excessive platelet activation in patients in the early, sub-acute, or late phases after transient ischemic attack or stroke in association with moderate or severe carotid stenosis and in asymptomatic moderate or severe carotid stenosis compared with controls. Furthermore, pilot data suggest that platelet activation may be increased in recently symptomatic than in asymptomatic severe carotid stenosis. Excessive platelet activation and platelet hyperreactivity may play a role in the pathogenesis of first or subsequent transient ischemic attack or stroke in patients with moderate or severe carotid stenosis. Larger longitudinal studies assessing platelet activation status with flow cytometry and platelet function\\/reactivity in symptomatic vs. asymptomatic carotid stenosis are warranted to improve our understanding of the mechanisms responsible for transient ischemic attack or stroke.

Nitrenes, carbenes, diradicals, and ylides. Interconversions of reactive intermediates.

Science.gov (United States)

Wentrup, Curt

2011-06-21

Rearrangements of aromatic and heteroaromatic nitrenes and carbenes can be initiated with either heat or light. The thermal reaction is typically induced by flash vacuum thermolysis, with isolation of the products at low temperatures. Photochemical experiments are conducted either under matrix isolation conditions or in solution at ambient temperature. These rearrangements are usually initiated by ring expansion of the nitrene or carbene to a seven-membered ring ketenimine, carbodiimide, or allene (that is, a cycloheptatetraene or an azacycloheptatetraene when a nitrogen is involved). Over the last few years, we have found that two types of ring opening take place as well. Type I is an ylidic ring opening that yields nitrile ylides or diazo compounds as transient intermediates. Type II ring opening produces either dienylnitrenes (for example, from 2-pyridylnitrenes) or 1,7-(1,5)-diradicals (such as those formed from 2-quinoxalinylnitrenes), depending on which of these species is better stabilized by resonance. In this Account, we describe our achievements in elucidating the nature of the ring-opened species and unraveling the connections between the various reactive intermediates. Both of these ring-opening reactions are found, at least in some cases, to dominate the subsequent chemistry. Examples include the formation of ring-opened ketenimines and carbodiimides, as well as the ring contraction reactions that form five-membered ring nitriles (such as 2- and 3-cyanopyrroles from pyridylnitrenes, N-cyanoimidazoles from 2-pyrazinyl and 4-pyrimidinylnitrenes, N-cyanopyrazoles from 2-pyrimidinylnitrenes and 3-pyridazinylnitrenes, and so forth). The mechanisms of formation of the open-chain and ring-contraction products were unknown at the onset of this study. In the course of our investigation, several reactions with three or more consecutive reactive intermediates have been unraveled, such as nitrene, seven-membered cyclic carbodiimide, and open-chain nitrile ylide

Preemptive, but not reactive, spinal cord stimulation mitigates transient ischemia-induced myocardial infarction via cardiac adrenergic neurons

NARCIS (Netherlands)

Southerland, E. M.; Milhorn, D. M.; Foreman, R. D.; Linderoth, B.; DeJongste, M. J. L.; Armour, J. A.; Subramanian, V.; Singh, M.; Singh, K.; Ardell, J. L.

2007-01-01

Our objective was to determine whether electrical neuromodulation using spinal cord stimulation ( SCS) mitigates transient ischemia-induced ventricular infarction and, if so, whether adrenergic neurons are involved in such cardioprotection. The hearts of anesthetized rabbits, subjected to 30 min of

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

Reactive dispersive contaminant transport in coastal aquifers: Numerical simulation of a reactive Henry problem

KAUST Repository

Nick, H.M.

2013-02-01

The reactive mixing between seawater and terrestrial water in coastal aquifers influences the water quality of submarine groundwater discharge. While these waters come into contact at the seawater groundwater interface by density driven flow, their chemical components dilute and react through dispersion. A larger interface and wider mixing zone may provide favorable conditions for the natural attenuation of contaminant plumes. It has been claimed that the extent of this mixing is controlled by both, porous media properties and flow conditions. In this study, the interplay between dispersion and reactive processes in coastal aquifers is investigated by means of numerical experiments. Particularly, the impact of dispersion coefficients, the velocity field induced by density driven flow and chemical component reactivities on reactive transport in such aquifers is studied. To do this, a hybrid finite-element finite-volume method and a reactive simulator are coupled, and model accuracy and applicability are assessed. A simple redox reaction is considered to describe the degradation of a contaminant which requires mixing of the contaminated groundwater and the seawater containing the terminal electron acceptor. The resulting degradation is observed for different scenarios considering different magnitudes of dispersion and chemical reactivity. Three reactive transport regimes are found: reaction controlled, reaction-dispersion controlled and dispersion controlled. Computational results suggest that the chemical components\\' reactivity as well as dispersion coefficients play a significant role on controlling reactive mixing zones and extent of contaminant removal in coastal aquifers. Further, our results confirm that the dilution index is a better alternative to the second central spatial moment of a plume to describe the mixing of reactive solutes in coastal aquifers. © 2012 Elsevier B.V.

Transient splenial lesion: Further experience with two cases

International Nuclear Information System (INIS)

Singh, Paramjeet; Gogoi, Dhrubajyoti; Vyas, Sameer; Khandelwal, Niranjan

2010-01-01

Transient splenial lesions (TSL) of the corpus callosum are uncommon radiologic findings that are seen in a number of clinical conditions with varied etiologies. They were first described a decade earlier in patients with epilepsy and hence were thought to be seizure or seizure therapy related. Subsequently, more cases were described by different observers in diseases with different etiologies, and the list is still increasing. Awareness of these lesions is necessary as they are an uncommon finding and have to be differentiated from other infective/noninfective causes. MRI is the imaging modality of choice as these lesions are not seen on routine noncontrast CT scan. The authors here describe two cases which showed TSL, with complete/partial resolution on follow-up scans. The authors also present a review of the literature

Transient splenial lesion: Further experience with two cases

Directory of Open Access Journals (Sweden)

Singh Paramjeet

2010-01-01

Full Text Available Transient splenial lesions (TSL of the corpus callosum are uncommon radiologic findings that are seen in a number of clinical conditions with varied etiologies. They were first described a decade earlier in patients with epilepsy and hence were thought to be seizure or seizure therapy related. Subsequently, more cases were described by different observers in diseases with different etiologies, and the list is still increasing. Awareness of these lesions is necessary as they are an uncommon finding and have to be differentiated from other infective/noninfective causes. MRI is the imaging modality of choice as these lesions are not seen on routine noncontrast CT scan. The authors here describe two cases which showed TSL, with complete/partial resolution on follow-up scans. The authors also present a review of the literature.

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.

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

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

Reactive Halogens in the Marine Boundary Layer (RHaMBLe: the tropical North Atlantic experiments

Directory of Open Access Journals (Sweden)

J. D. Lee

2010-02-01

Full Text Available The NERC UK SOLAS-funded Reactive Halogens in the Marine Boundary Layer (RHaMBLe programme comprised three field experiments. This manuscript presents an overview of the measurements made within the two simultaneous remote experiments conducted in the tropical North Atlantic in May and June 2007. Measurements were made from two mobile and one ground-based platforms. The heavily instrumented cruise D319 on the RRS Discovery from Lisbon, Portugal to São Vicente, Cape Verde and back to Falmouth, UK was used to characterise the spatial distribution of boundary layer components likely to play a role in reactive halogen chemistry. Measurements onboard the ARSF Dornier aircraft were used to allow the observations to be interpreted in the context of their vertical distribution and to confirm the interpretation of atmospheric structure in the vicinity of the Cape Verde islands. Long-term ground-based measurements at the Cape Verde Atmospheric Observatory (CVAO on São Vicente were supplemented by long-term measurements of reactive halogen species and characterisation of additional trace gas and aerosol species during the intensive experimental period.

This paper presents a summary of the measurements made within the RHaMBLe remote experiments and discusses them in their meteorological and chemical context as determined from these three platforms and from additional meteorological analyses. Air always arrived at the CVAO from the North East with a range of air mass origins (European, Atlantic and North American continental. Trace gases were present at stable and fairly low concentrations with the exception of a slight increase in some anthropogenic components in air of North American origin, though NO_x mixing ratios during this period remained below 20 pptv (note the non-IUPAC adoption in this manuscript of pptv and ppbv, equivalent to pmol mol⁻¹ and nmol mol⁻¹ to reflect common practice. Consistency with

Enhanced Severe Transient Analysis for Prevention Technical Program Plan

Energy Technology Data Exchange (ETDEWEB)

Gougar, Hans [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s major emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code

Some safety related characteristics of Phenix, a 250 MWe fast reactor -1989 and 1990 negative reactivity trip investigations

International Nuclear Information System (INIS)

Chaumont, J.M.; Goux, D.; Martin, L.

1993-01-01

The main characteristics of the Phenix core control are summarized. The current state of the investigations related to the 1989 and 1990 negative reactivity transients are presented with emphasis on the results of the very low power tests recently performed. (authors). 5 figs., 2 refs

Preliminary investigation of actinide and xenon reactivity effects in accelerator transmutation of waste high-flux systems

International Nuclear Information System (INIS)

Olson, K.R.; Henderson, D.L.

1995-01-01

The possibility of an unstable positive reactivity growth in an accelerator transmutation of waste (ATW)-type high-flux system is investigated. While it has always been clear that xenon is an important actor in the reactivity response of a system to flux changes, it has been suggested that in very high thermal flux transuranic burning systems, a positive, unstable reactivity growth could be caused by the actinides alone. Initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately. The maximum change in reactivity after a flux change caused by the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or startup. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response caused by the actinides. The capabilities and applications of both the current actinide model and the xenon model are discussed. Finally, the need for a complete dynamic model for the high-flux fluid-fueled ATW system is addressed

The excess radio background and fast radio transients

International Nuclear Information System (INIS)

Kehayias, John; Kephart, Thomas W.; Weiler, Thomas J.

2015-01-01

In the last few years ARCADE 2, combined with older experiments, has detected an additional radio background, measured as a temperature and ranging in frequency from 22 MHz to 10 GHz, not accounted for by known radio sources and the cosmic microwave background. One type of source which has not been considered in the radio background is that of fast transients (those with event times much less than the observing time). We present a simple estimate, and a more detailed calculation, for the contribution of radio transients to the diffuse background. As a timely example, we estimate the contribution from the recently-discovered fast radio bursts (FRBs). Although their contribution is likely 6 or 7 orders of magnitude too small (though there are large uncertainties in FRB parameters) to account for the ARCADE 2 excess, our development is general and so can be applied to any fast transient sources, discovered or yet to be discovered. We estimate parameter values necessary for transient sources to noticeably contribute to the radio background

Assessment of the turbine trip transient in Cofrentes NPP with TRAC-BF1

International Nuclear Information System (INIS)

Castrillo, F.; Gomez, A.; Gallego, I.

1993-06-01

This report presents the results of the assessment of TRAC-BF1 (G1-J1) code with the model of C. N. Cofrentes for simulation of the transient originated by the manual trip of the main turbine. C. N. Cofrentes is a General Electric designed BWR/6 plant, with a nominal core thermal power of 2894 Mwt, in commercial operation since 1985, owned and operated by Hidroelectrica Espanola, S. A. The plant incorporates all the characteristics of BWR/6 reactors, with two turbine driven FW pumps. As a result of this assessment a model of C. N. Cofrentes has been developed for TRAC-BF1 that fairly reproduces operational transient behavior of the plant. A special purpose code was generated to obtain reactivity coefficients, as required by TRAC-BF1, from the 3D simulator

Borylnitrenes: electrophilic reactive intermediates with high reactivity towards C-H bonds.

Science.gov (United States)

Bettinger, Holger F; Filthaus, Matthias

2010-12-21

Borylnitrenes (catBN 3a and pinBN 3b; cat = catecholato, pin = pinacolato) are reactive intermediates that show high tendency towards insertion into the C-H bonds of unactivated hydrocarbons. The present article summarizes the matrix isolation investigations that were aimed at identifying, characterizing and investigating the chemical behaviour of 3a by spectroscopic means, and of the experiments in solution and in the gas phase that were performed with 3b. Comparison with the reactivity reported for difluorovinylidene 1a in solid argon indicates that 3a shows by and large similar reactivity, but only after photochemical excitation. The derivative 3b inserts into the C-H bonds of hydrocarbon solvents in high yields and thus allows the formation of primary amines, secondary amines, or amides from "unreactive" hydrocarbons. It can also be used for generation of methylamine or methylamide from methane in the gas phase at room temperature. Remaining challenges in the chemistry of borylnitrenes are briefly summarized.

Referencing cross-reactivity of detection antibodies for protein array experiments [version 1; referees: 1 approved, 2 approved with reservations

Directory of Open Access Journals (Sweden)

Darragh Lemass

2016-01-01

Full Text Available Protein arrays are frequently used to profile antibody repertoires in humans and animals. High-throughput protein array characterisation of complex antibody repertoires requires a platform-dependent, lot-to-lot validation of secondary detection antibodies. This article details the validation of an affinity-isolated anti-chicken IgY antibody produced in rabbit and a goat anti-rabbit IgG antibody conjugated with alkaline phosphatase using protein arrays consisting of 7,390 distinct human proteins. Probing protein arrays with secondary antibodies in absence of chicken serum revealed non-specific binding to 61 distinct human proteins. The cross-reactivity of the tested secondary detection antibodies points towards the necessity of platform-specific antibody characterisation studies for all secondary immunoreagents. Secondary antibody characterisation using protein arrays enables generation of reference lists of cross-reactive proteins, which can be then excluded from analysis in follow-up experiments. Furthermore, making such cross-reactivity lists accessible to the wider research community may help to interpret data generated by the same antibodies in applications not related to protein arrays such as immunoprecipitation, Western blots or other immunoassays.

Sensitivity of reactivity feedback due to core bowing in a metallic-fueled core

International Nuclear Information System (INIS)

Nakagawa, Masatoshi; Kawashima, Masatoshi; Endo, Hiroshi; Nishimura, Tomohiro

1991-01-01

A sensitivity study has been carried out on negative reactivity feedback caused by core bowing to assess the potential effectiveness of FBR passive safety features in regard to withstanding an anticipated transient without scram (ATWS). In the present study, an analysis has been carried to obtain the best material and geometrical conditions concerning the core restraint system out for several power to flow rates (P/F), up to 2.0 for a 300 MWe metallic-fueled core. From this study, it was clarified that the pad stiffness at an above core loading pads (ACLP) needs to be large enough to ensure negative reactivity feedback against ATWS. It was also clarified that there is an upper limit for the clearances between ducts at ACLP. A new concept, in regard to increasing the absolute value for negative reactivity feedback due to core bowing at ATWS, is proposed and discussed. (author)

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.

Characterization of transient gain x-ray lasers

International Nuclear Information System (INIS)

Dunn, J.; Osterheld, A.; Shlyaptsev, V.

1999-01-01

We have performed numerical simulations of the transient collisional excitation Ni-like Pd 4d → 4p J = 0 → 1 147 angstrom laser transition recently observed at Lawrence Livermore National Laboratory (LLNL). The high gain ∼35 cm results from the experiment are compared with detailed modeling simulations from the 1-D RADEX code in order to better understand the main physics issues affecting the measured gain and x-ray laser propagation along the plasma column. Simulations indicate that the transient gain lifetime associated with the short pulse pumping and refraction of the x-ray laser beam out of the gain region are the main detrimental effects. Gain lifetimes of ∼7 ps(1/e decay) are inferred from the smoothly changing gain experimental observations and are in good agreement with the simulations. Furthermore, the modeling results indicate the presence of a longer-lived but lower gain later in time associated with the transition from transient to quasi-steady state excitation

Experimental demonstration of tokamak inductive flux saving by transient coaxial helicity injection on national spherical torus experiment

Energy Technology Data Exchange (ETDEWEB)

Raman, R.; Jarboe, T. R.; Nelson, B. A. [University of Washington, Seattle, Washington 98195 (United States); Mueller, D.; Bell, M. G.; Gerhardt, S.; LeBlanc, B.; Menard, J.; Ono, M.; Roquemore, L. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

2011-09-15

Discharges initiated by transient coaxial helicity injection in National Spherical Torus Experiment have attained peak toroidal plasma currents up to 300 kA. When induction from the central solenoid is then applied, these discharges develop up to 300 kA additional current compared to discharges initiated by induction only. CHI initiated discharges in NSTX have achieved 1 MA of plasma current using only 258 mWb of solenoid flux whereas standard induction-only discharges require about 50% more solenoid flux to reach 1 MA. In addition, the CHI-initiated discharge has lower plasma density and a low normalized internal plasma inductance of 0.35, as needed for achieving advanced scenarios in NSTX.

Photodetachment and electron reactivity in 1-methyl-1-butyl-pyrrolidinium bis(trifluoromethylsulfonyl)amide

Energy Technology Data Exchange (ETDEWEB)

Molins i Domenech, Francesc; FitzPatrick, Benjamin; Healy, Andrew T.; Blank, David A. [Department of Chemistry, University of Minnesota, 207 Pleasant St. SE, Minneapolis, Minnesota 55455 (United States)

2012-07-21

The transient absorption spectrum in the range 500 nm-1000 nm was measured with ultrafast time resolution on a flowing neat, aliphatic, room-temperature ionic liquid following anion photodetachment. In this region the spectrum was shown to be a combination of absorption from the electron and the hole. Spectrally-resolved electron quenching determined a bimodal shape for the hole spectrum in agreement with recent computational predictions on a smaller aliphatic ionic liquid [Margulis et al., J. Am. Chem. Soc. 133, 20186 (2011)]. For time delays beyond 15 ps, spectral evolution qualitatively agrees with recent radiolysis experiments [Wishart et al., Faraday Discuss. 154, 353 (2012)]. However, the shape of the spectrum is different, reflecting the contrast in ionization energy between the two methods. Previously unobserved reactivity of the electron was found with a time constant of 300 fs. The results demonstrate solvent control of the rate coefficient for reaction between the electron and proton, with a rapid decline in the rate within the first picosecond.

The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

International Nuclear Information System (INIS)

Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

2006-01-01

The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

Transient Mathematical Modeling for Liquid Rocket Engine Systems: Methods, Capabilities, and Experience

Science.gov (United States)

Seymour, David C.; Martin, Michael A.; Nguyen, Huy H.; Greene, William D.

2005-01-01

The subject of mathematical modeling of the transient operation of liquid rocket engines is presented in overview form from the perspective of engineers working at the NASA Marshall Space Flight Center. The necessity of creating and utilizing accurate mathematical models as part of liquid rocket engine development process has become well established and is likely to increase in importance in the future. The issues of design considerations for transient operation, development testing, and failure scenario simulation are discussed. An overview of the derivation of the basic governing equations is presented along with a discussion of computational and numerical issues associated with the implementation of these equations in computer codes. Also, work in the field of generating usable fluid property tables is presented along with an overview of efforts to be undertaken in the future to improve the tools use for the mathematical modeling process.

Insights into the reactivation of cobalamin-dependent methionine synthase

Energy Technology Data Exchange (ETDEWEB)

Koutmos, Markos; Datta, Supratim; Pattridge, Katherine A.; Smith, Janet L.; Matthews, Rowena G.; (Michigan)

2009-12-10

Cobalamin-dependent methionine synthase (MetH) is a modular protein that catalyzes the transfer of a methyl group from methyltetrahydrofolate to homocysteine to produce methionine and tetrahydrofolate. The cobalamin cofactor, which serves as both acceptor and donor of the methyl group, is oxidized once every {approx}2,000 catalytic cycles and must be reactivated by the uptake of an electron from reduced flavodoxin and a methyl group from S-adenosyl-L-methionine (AdoMet). Previous structures of a C-terminal fragment of MetH (MetH{sup CT}) revealed a reactivation conformation that juxtaposes the cobalamin- and AdoMet-binding domains. Here we describe 2 structures of a disulfide stabilized MetH{sup CT} ({sub s-s}MetH{sup CT}) that offer further insight into the reactivation of MetH. The structure of {sub s-s}MetH{sup CT} with cob(II)alamin and S-adenosyl-L-homocysteine represents the enzyme in the reactivation step preceding electron transfer from flavodoxin. The structure supports earlier suggestions that the enzyme acts to lower the reduction potential of the Co(II)/Co(I) couple by elongating the bond between the cobalt and its upper axial water ligand, effectively making the cobalt 4-coordinate, and illuminates the role of Tyr-1139 in the stabilization of this 4-coordinate state. The structure of {sub s-s}MetH{sub CT} with aquocobalamin may represent a transient state at the end of reactivation as the newly remethylated 5-coordinate methylcobalamin returns to the 6-coordinate state, triggering the rearrangement to a catalytic conformation.

5-Azacytidine mediated reactivation of silenced transgenes in potato (Solanum tuberosum) at the whole plant level.

Science.gov (United States)

Tyč, Dimitrij; Nocarová, Eva; Sikorová, Lenka; Fischer, Lukáš

2017-08-01

Transient 5-azacytidine treatment of leaf explants from potato plants with transcriptionally silenced transgenes allows de novo regeneration of plants with restored transgene expression at the whole plant level. Transgenes introduced into plant genomes frequently become silenced either at the transcriptional or the posttranscriptional level. Transcriptional silencing is usually associated with DNA methylation in the promoter region. Treatments with inhibitors of maintenance DNA methylation were previously shown to allow reactivation of transcriptionally silenced transgenes in single cells or tissues, but not at the whole plant level. Here we analyzed the effect of DNA methylation inhibitor 5-azacytidine (AzaC) on the expression of two silenced reporter genes encoding green fluorescent protein (GFP) and neomycin phosphotransferase (NPTII) in potato plants. Whereas no obvious reactivation was observed in AzaC-treated stem cuttings, transient treatment of leaf segments with 10 μM AzaC and subsequent de novo regeneration of shoots on the selective medium with kanamycin resulted in the production of whole plants with clearly reactivated expression of previously silenced transgenes. Reactivation of nptII expression was accompanied by a decrease in cytosine methylation in the promoter region of the gene. Using the plants with reactivated GFP expression, we found that re-silencing of this transgene can be accidentally triggered by de novo regeneration. Thus, testing the incidence of transgene silencing during de novo regeneration could be a suitable procedure for negative selection of transgenic lines (insertion events) which have an inclination to be silenced. Based on our analysis of non-specific inhibitory effects of AzaC on growth of potato shoots in vitro, we estimated that AzaC half-life in the culture media is approximately 2 days.

A novel dual-wavelength laser stimulator to elicit transient and tonic nociceptive stimulation.

Science.gov (United States)

Dong, Xiaoxi; Liu, Tianjun; Wang, Han; Yang, Jichun; Chen, Zhuying; Hu, Yong; Li, Yingxin

2017-07-01

This study aimed to develop a new laser stimulator to elicit both transient and sustained heat stimulation with a dual-wavelength laser system as a tool for the investigation of both transient and tonic experimental models of pain. The laser stimulator used a 980-nm pulsed laser to generate transient heat stimulation and a 1940-nm continuous-wave (CW) laser to provide sustained heat stimulation. The laser with 980-nm wavelength can elicit transient pain with less thermal injury, while the 1940-nm CW laser can effectively stimulate both superficial and deep nociceptors to elicit tonic pain. A proportional integral-derivative (PID) temperature feedback control system was implemented to ensure constancy of temperature during heat stimulation. The performance of this stimulator was evaluated by in vitro and in vivo animal experiments. In vitro experiments on totally 120 specimens fresh pig skin included transient heat stimulation by 980-nm laser (1.5 J, 10 ms), sustained heat stimulation by 1940-nm laser (50-55 °C temperature control mode or 1.5 W, 5 min continuous power supply), and the combination of transient/sustained heat stimulation by dual lasers (1.5 J, 10 ms, 980-nm pulse laser, and 1940-nm laser with 50-55 °C temperature control mode). Hemoglobin brushing and wind-cooling methods were tested to find better stimulation model. A classic tail-flick latency (TFL) experiment with 20 Wistar rats was used to evaluate the in vivo efficacy of transient and tonic pain stimulation with 15 J, 100 ms 980-nm single laser pulse, and 1.5 W constant 1940-nm laser power. Ideal stimulation parameters to generate transient pain were found to be a 26.6 °C peak temperature rise and 0.67 s pain duration. In our model of tonic pain, 5 min of tonic stimulation produced a temperature change of 53.7 ± 1.3 °C with 1.6 ± 0.2% variation. When the transient and tonic stimulation protocols were combined, no significant difference was observed depending on the order

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

Use of plant operating history to define transient loads

International Nuclear Information System (INIS)

Dwivedy, K.K.

1996-01-01

Fatigue and crack growth analyses of components subjected to transient loads have been under continuous development during the recent past to include effects of environment on the components. The accuracy of the evaluation method on the predicted reliability of the components in the operating environment has become a focus of attention. Methods have integrated available material/component test data to improve evaluation techniques. However, in the area of definition of thermal transient loads the analyst still has to remain conservative, because no realistic guidelines have been developed to define thermal transients and their sequences. Fatigue re-evaluations of components are becoming increasingly necessary in operating plants as they age due to two reasons: (1) Components show age related degradation and cannot be repaired/replaced due to economic/logistic reasons. (2) Components experience transient conditions which were not considered in the original design. In either case, the evaluation of remaining life of components involves definition of transients and their sequence from the time the component was put in service until the end of life. As a common practice, initial plant design transients are used in a conservative definition of sequences to obtain results unrealistic for the situation, which sometimes leads to inaccurate estimate of the remaining life of components. The objective of this paper is to use plant operating history and plant monitoring data to provide procedures and techniques to define realistic transients for evaluation

The influence of spatial effects on the measurement results of reactivity in 'fast disturbances' of core parameters

International Nuclear Information System (INIS)

Tsyganov, S.V.; Shishkov, L.K.

2001-01-01

The analysis of methods for the determination of reactivity revealed an essential influence of spatial effect on the measurement precision. Using of reverse point kinetic equation for reactivity meter is assumed that the average neutron flux weigh with the importance function is known at every moment of the transient. In fact, reactivity meter represent behaviour of the neutron flux only of the part of the core, so measured value of reactivity can differ from really reactivity. Three-dimensional dynamic model of the core allow to evaluate such difference. It is supposed to evaluate correction factor for the neutron flux measured at the place where ion chamber situated with the three-dimensional model NOSTRA of the WWER core. On the basis of such algorithm we propose to build module allowing the influence of spatial effects on the results of the reactivity meter to be eliminated at real time regime. This code will be incorporated into the core monitoring system 'BLOK' (SCORPIO type) which is being developed for the Kola and Rostov NPP. The report illustrates utilization of such algorithm (Authors)

Rap1 signaling is required for suppression of Ras-generated reactive oxygen species and protection against oxidative stress in T lymphocytes

NARCIS (Netherlands)

Remans, Philip H. J.; Gringhuis, Sonja I.; van Laar, Jacob M.; Sanders, Marjolein E.; Papendrecht-van der Voort, Ellen A. M.; Zwartkruis, Fried J. T.; Levarht, E. W. Nivine; Rosas, Marcela; Coffer, Paul J.; Breedveld, Ferdinand C.; Bos, Johannes L.; Tak, Paul P.; Verweij, Cornelis L.; Reedquist, Kris A.

2004-01-01

Transient production of reactive oxygen species (ROS) plays an important role in optimizing transcriptional and proliferative responses to TCR signaling in T lymphocytes. Conversely, chronic oxidative stress leads to decreased proliferative responses and enhanced transcription of inflammatory gene

DETECTION OF FAST TRANSIENTS WITH RADIO INTERFEROMETRIC ARRAYS

International Nuclear Information System (INIS)

Bhat, N. D. R.; Chengalur, J. N.; Gupta, Y.; Prasad, J.; Roy, J.; Kudale, S. S.; Cox, P. J.; Bailes, M.; Burke-Spolaor, S.; Van Straten, W.

2013-01-01

Next-generation radio arrays, including the Square Kilometre Array (SKA) and its pathfinders, will open up new avenues for exciting transient science at radio wavelengths. Their innovative designs, comprising a large number of small elements, pose several challenges in digital processing and optimal observing strategies. The Giant Metre-wave Radio Telescope (GMRT) presents an excellent test-bed for developing and validating suitable observing modes and strategies for transient experiments with future arrays. Here we describe the first phase of the ongoing development of a transient detection system for GMRT that is planned to eventually function in a commensal mode with other observing programs. It capitalizes on the GMRT's interferometric and sub-array capabilities, and the versatility of a new software backend. We outline considerations in the plan and design of transient exploration programs with interferometric arrays, and describe a pilot survey that was undertaken to aid in the development of algorithms and associated analysis software. This survey was conducted at 325 and 610 MHz, and covered 360 deg 2 of the sky with short dwell times. It provides large volumes of real data that can be used to test the efficacies of various algorithms and observing strategies applicable for transient detection. We present examples that illustrate the methodologies of detecting short-duration transients, including the use of sub-arrays for higher resilience to spurious events of terrestrial origin, localization of candidate events via imaging, and the use of a phased array for improved signal detection and confirmation. In addition to demonstrating applications of interferometric arrays for fast transient exploration, our efforts mark important steps in the roadmap toward SKA-era science.

Detection of Fast Transients with Radio Interferometric Arrays

Science.gov (United States)

Bhat, N. D. R.; Chengalur, J. N.; Cox, P. J.; Gupta, Y.; Prasad, J.; Roy, J.; Bailes, M.; Burke-Spolaor, S.; Kudale, S. S.; van Straten, W.

2013-05-01

Next-generation radio arrays, including the Square Kilometre Array (SKA) and its pathfinders, will open up new avenues for exciting transient science at radio wavelengths. Their innovative designs, comprising a large number of small elements, pose several challenges in digital processing and optimal observing strategies. The Giant Metre-wave Radio Telescope (GMRT) presents an excellent test-bed for developing and validating suitable observing modes and strategies for transient experiments with future arrays. Here we describe the first phase of the ongoing development of a transient detection system for GMRT that is planned to eventually function in a commensal mode with other observing programs. It capitalizes on the GMRT's interferometric and sub-array capabilities, and the versatility of a new software backend. We outline considerations in the plan and design of transient exploration programs with interferometric arrays, and describe a pilot survey that was undertaken to aid in the development of algorithms and associated analysis software. This survey was conducted at 325 and 610 MHz, and covered 360 deg2 of the sky with short dwell times. It provides large volumes of real data that can be used to test the efficacies of various algorithms and observing strategies applicable for transient detection. We present examples that illustrate the methodologies of detecting short-duration transients, including the use of sub-arrays for higher resilience to spurious events of terrestrial origin, localization of candidate events via imaging, and the use of a phased array for improved signal detection and confirmation. In addition to demonstrating applications of interferometric arrays for fast transient exploration, our efforts mark important steps in the roadmap toward SKA-era science.

Analysis of transients for NPP with VVER-440 using the code SiTAP

International Nuclear Information System (INIS)

Kalinenko, V.

1994-06-01

The report contains analysis of transients ''Loop connection'' and ''Steam generator tube rupture'' for nuclear power plants (NPP) with VVER-440. To obtain more detailed information about NPP's dynamic characteristics, various variants of initial and boundary conditions are considerd. Calculation of these transients was performed using the SiTAP code developed at the Nuclear Safety Institute of the Russian Research Centre ''Kurchatov Institute''. SiTAP code is a multifunctional computer tool for fast analysis of transient and accidental processes of VVER type reactors for engineers working in the field of NPP dynamics. SiTAP can be used form comparative analysis of several variants of accident scenarios to find out the conditions leading to most serious consequences from a safety point of view. In such cases, additional analyses using best-estimate codes should be carried out. The results of SiTAP for a faulty loop connection leading to a boron dilution accident are intended to be used as boundary conditions for a more detailed anlaysis with the aid of the three-dimensional reactor core model DYN3D, developed in the Research Centre Rossendorf for the simulation of reactivity initiated accidents. (orig.)

Quantifying and Predicting Three-Dimensional Heterogeneity in Transient Storage Using Roving Profiling

Science.gov (United States)

Kaplan, D. A.; Reaver, N.; Hensley, R. T.; Cohen, M. J.

2017-12-01

Hydraulic transport is an important component of nutrient spiraling in streams. Quantifying conservative solute transport is a prerequisite for understanding the cycling and fate of reactive solutes, such as nutrients. Numerous studies have modeled solute transport within streams using the one-dimensional advection, dispersion and storage (ADS) equation calibrated to experimental data from tracer experiments. However, there are limitations to the information about in-stream transient storage that can be derived from calibrated ADS model parameters. Transient storage (TS) in the ADS model is most often modeled as a single process, and calibrated model parameters are "lumped" values that are the best-fit representation of multiple real-world TS processes. In this study, we developed a roving profiling method to assess and predict spatial heterogeneity of in-stream TS. We performed five tracer experiments on three spring-fed rivers in Florida (USA) using Rhodamine WT. During each tracer release, stationary fluorometers were deployed to measure breakthrough curves for multiple reaches within the river. Teams of roving samplers moved along the rivers measuring tracer concentrations at various locations and depths within the reaches. A Bayesian statistical method was used to calibrate the ADS model to the stationary breakthrough curves, resulting in probability distributions for both the advective and TS zone as a function of river distance and time. Rover samples were then assigned a probability of being from either the advective or TS zone by comparing measured concentrations to the probability distributions of concentrations in the ADS advective and TS zones. A regression model was used to predict the probability of any in-stream position being located within the advective versus TS zone based on spatiotemporal predictors (time, river position, depth, and distance from bank) and eco-geomorphological feature (eddies, woody debris, benthic depressions, and aquatic

Measurement and analysis of reactivity temperature coefficient of CEFR

International Nuclear Information System (INIS)

Chen Yiyu; Hu Yun; Yang Xiaoyan; Fan Zhendong; Zhang Qiang; Zhao Jinkun; Li Zehua

2013-01-01

The reactivity temperature coefficient of CEFR was calculated by CITATION program and compared with the results calculated by correlative programs and measured from experiments for temperature effects. It is indicated that the calculation results from CITATION agree well with measured values. The reactivity temperature coefficient of CEFR is about -4 pcm/℃. The deviation of the measured values between the temperature increasing and decreasing processes is about 11%, which satisfies the experiment acceptance criteria. The measured results can validate the calculation ones by program and can provide important reference data for the safety operation of CEFR and the analysis of the reactivity balance in the reactor refueling situation. (authors)

Current interruption transients calculation

CERN Document Server

Peelo, David F

2014-01-01

Provides an original, detailed and practical description of current interruption transients, origins, and the circuits involved, and how they can be calculated Current Interruption Transients Calculationis a comprehensive resource for the understanding, calculation and analysis of the transient recovery voltages (TRVs) and related re-ignition or re-striking transients associated with fault current interruption and the switching of inductive and capacitive load currents in circuits. This book provides an original, detailed and practical description of current interruption transients, origins,

The perturbation theory in the fundamental mode. Its application to the analysis of neutronic experiments involving small amounts of materials in fast neutron multiplying media

International Nuclear Information System (INIS)

Remsak, Stanislav.

1975-01-01

The formalism of the perturbation theory at the first order, is developed in its simplest form: diffusion theory in the fundamental mode and then the more complex formalism of the transport theory in the fundamental mode. A comparison shows the effect of the angular correlation between the fine structures of the flux and its adjoint function, the difference in the treatment of neutron leakage phenomena, and the existence of new terms in the perturbation formula, entailing a reactivity representation in the diffusion theory that is not quite exact. Problems of using the formalism developed are considered: application of the multigroup formalism, transients of the flux and its adjoint function, validity of the first order approximation etc. A detailed analysis allows the formulation of a criterion specifying the validity range. Transients occuring in the reference medium are also treated. A set of numerical tests for determining a method of elimination of transient effects is presented. Some differential experiments are then discussed: sodium blowdown in enriched uranium or plutonium cores, experiments utilizing some structural materials (iron and oxygen) and plutonium sample oscillations. The Cadarache version II program was systematically used but the analysis of the experiments of plutonium sample oscillation in Ermine required the Cadarache version III program [fr

Parametric study of a reactivity accident in a pressurized water reactor: control rod cluster ejection

International Nuclear Information System (INIS)

Chesnel, A.

1985-01-01

This research thesis concerns a class 4 accident in a PWR: the ejection of a control rod cluster from the reactor core. It aims at defining, for such an accident, the envelope values which relate the reactivity to the hot spot factor within the frame of a mode A control. The report describes the physical phenomena and their modelling during the considered transient. It presents a simple mathematical solution of the accident which shows that the main neutron parameters are the released reactivity, the delayed neutron fraction, the Doppler coefficient, and the hot spot factor. It reports a temperature sensitivity study, and discusses three-dimensional calculations of irradiation distributions

Stress reactivity and personality in extreme sport athletes: The psychobiology of BASE jumpers.

Science.gov (United States)

Monasterio, Erik; Mei-Dan, Omer; Hackney, Anthony C; Lane, Amy R; Zwir, Igor; Rozsa, Sandor; Cloninger, C Robert

2016-12-01

This is the first report of the psychobiology of stress in BASE jumpers, one of the most dangerous forms of extreme sport. We tested the hypotheses that indicators of emotional style (temperament) predict salivary cortisol reactivity, whereas indicators of intentional goal-setting (persistence and character) predict salivary alpha-amylase reactivity during BASE jumping. Ninety-eight subjects completed the Temperament and Character Inventory (TCI) the day before the jump, and 77 also gave salivary samples at baseline, pre-jump on the bridge over the New River Gorge, and post-jump upon landing. Overall BASE jumpers are highly resilient individuals who are highly self-directed, persistent, and risk-taking, but they are heterogeneous in their motives and stress reactivity in the Hypothalamic-Pituitary-Adrenal (HPA) stress system (cortisol reactivity) and the sympathetic arousal system (alpha-amylase reactivity). Three classes of jumpers were identified using latent class analysis based on their personality profiles, prior jumping experience, and levels of cortisol and alpha-amylase at all three time points. "Masterful" jumpers (class 1) had a strong sense of self-directedness and mastery, extensive prior experience, and had little alpha-amylase reactivity and average cortisol reactivity. "Trustful" jumpers (class 2) were highly cooperative and trustful individuals who had little cortisol reactivity coincident with the social support they experienced prior to jumping. "Courageous" jumpers (class 3) were determined despite anxiety and inexperience, and they had high sympathetic reactivity but average cortisol activation. We conclude that trusting social attachment (Reward Dependence) and not jumping experience predicted low cortisol reactivity, whereas persistence (determination) and not jumping experience predicted high alpha-amylase reactivity. Copyright © 2016 The Authors. Published by Elsevier Inc. All rights reserved.

The development of the fuel rod transient performance analysis code FTPAC

International Nuclear Information System (INIS)

Han Zhijie; Ji Songtao

2014-01-01

Fuel rod behavior, especially the integrity of cladding, played an important role in fuel safety research during reactor transient and hypothetical accidents conditions. In order to study fuel rod performance under transient accidents, FTPAC (Fuel Transient Performance Analysis Code) has been developed for simulating light water reactor fuel rod transient behavior when power or coolant boundary conditions are rapidly changing. It is composed of temperature, mechanical deformation, cladding oxidation and gas pressure model. The assessment was performed by comparing FTPAC code analysis result to experiments data and FRAPTRAN code calculations. Comparison shows that, the FTPAC gives reasonable agreement in temperature, deformation and gas pressure prediction. And the application of slip coefficient is more suitable for simulating the sliding between pellet and cladding when the gap is closed. (authors)

porewater chemistry experiment at Mont Terri rock laboratory. Reactive transport modelling including bacterial activity

International Nuclear Information System (INIS)

Tournassat, Christophe; Gaucher, Eric C.; Leupin, Olivier X.; Wersin, Paul

2010-01-01

Document available in extended abstract form only. An in-situ test in the Opalinus Clay formation, termed pore water Chemistry (PC) experiment, was run for a period of five years. It was based on the concept of diffusive equilibration whereby traced water with a composition close to that expected in the formation was continuously circulated and monitored in a packed off borehole. The main original focus was to obtain reliable data on the pH/pCO 2 of the pore water, but because of unexpected microbially- induced redox reactions, the objective was then changed to elucidate the biogeochemical processes happening in the borehole and to understand their impact on pH/pCO 2 and pH in the low permeability clay formation. The biologically perturbed chemical evolution of the PC experiment was simulated with reactive transport models. The aim of this modelling exercise was to develop a 'minimal-' model able to reproduce the chemical evolution of the PC experiment, i.e. the chemical evolution of solute inorganic and organic compounds (organic carbon, dissolved inorganic carbon etc...) that are coupled with each other through the simultaneous occurrence of biological transformation of solute or solid compounds, in-diffusion and out-diffusion of solute species and precipitation/dissolution of minerals (in the borehole and in the formation). An accurate description of the initial chemical conditions in the surrounding formation together with simplified kinetics rule mimicking the different phases of bacterial activities allowed reproducing the evolution of all main measured parameters (e.g. pH, TOC). Analyses from the overcoring and these simulations evidence the high buffer capacity of Opalinus clay regarding chemical perturbations due to bacterial activity. This pH buffering capacity is mainly attributed to the carbonate system as well as to the clay surfaces reactivity. Glycerol leaching from the pH-electrode might be the primary organic source responsible for

Development of an alternative reactivity meter for nuclear reactor control

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1991-01-01

This work describes an alternative version of the IPEN-CNEN/SP reactivity-meter. This new version utilizes a programmable electrometer (to realize the data acquisition) and a IBM-PC microcomputer to process the reactivity calculation. The aim of development of this alternative reactivity-meter is to have available a equipment of measurements of reactivity in the case of the later version show any problem during an experiment. (author)

Reactivation of chromosomally integrated human herpesvirus-6 by telomeric circle formation.

Directory of Open Access Journals (Sweden)

Bhupesh K Prusty

Full Text Available More than 95% of the human population is infected with human herpesvirus-6 (HHV-6 during early childhood and maintains latent HHV-6 genomes either in an extra-chromosomal form or as a chromosomally integrated HHV-6 (ciHHV-6. In addition, approximately 1% of humans are born with an inheritable form of ciHHV-6 integrated into the telomeres of chromosomes. Immunosuppression and stress conditions can reactivate latent HHV-6 replication, which is associated with clinical complications and even death. We have previously shown that Chlamydia trachomatis infection reactivates ciHHV-6 and induces the formation of extra-chromosomal viral DNA in ciHHV-6 cells. Here, we propose a model and provide experimental evidence for the mechanism of ciHHV-6 reactivation. Infection with Chlamydia induced a transient shortening of telomeric ends, which subsequently led to increased telomeric circle (t-circle formation and incomplete reconstitution of circular viral genomes containing single viral direct repeat (DR. Correspondingly, short t-circles containing parts of the HHV-6 DR were detected in cells from individuals with genetically inherited ciHHV-6. Furthermore, telomere shortening induced in the absence of Chlamydia infection also caused circularization of ciHHV-6, supporting a t-circle based mechanism for ciHHV-6 reactivation.

Establishment of HSV1 latency in immunodeficient mice facilitates efficient in vivo reactivation.

Directory of Open Access Journals (Sweden)

Chandran Ramakrishna

2015-03-01

Full Text Available The establishment of latent infections in sensory neurons is a remarkably effective immune evasion strategy that accounts for the widespread dissemination of life long Herpes Simplex Virus type 1 (HSV1 infections in humans. Periodic reactivation of latent virus results in asymptomatic shedding and transmission of HSV1 or recurrent disease that is usually mild but can be severe. An in-depth understanding of the mechanisms regulating the maintenance of latency and reactivation are essential for developing new approaches to block reactivation. However, the lack of a reliable mouse model that supports efficient in vivo reactivation (IVR resulting in production of infectious HSV1 and/or disease has hampered progress. Since HSV1 reactivation is enhanced in immunosuppressed hosts, we exploited the antiviral and immunomodulatory activities of IVIG (intravenous immunoglobulins to promote survival of latently infected immunodeficient Rag mice. Latently infected Rag mice derived by high dose (HD, but not low dose (LD, HSV1 inoculation exhibited spontaneous reactivation. Following hyperthermia stress (HS, the majority of HD inoculated mice developed HSV1 encephalitis (HSE rapidly and synchronously, whereas for LD inoculated mice reactivated HSV1 persisted only transiently in trigeminal ganglia (Tg. T cells, but not B cells, were required to suppress spontaneous reactivation in HD inoculated latently infected mice. Transfer of HSV1 memory but not OVA specific or naïve T cells prior to HS blocked IVR, revealing the utility of this powerful Rag latency model for studying immune mechanisms involved in control of reactivation. Crossing Rag mice to various knockout strains and infecting them with wild type or mutant HSV1 strains is expected to provide novel insights into the role of specific cellular and viral genes in reactivation, thereby facilitating identification of new targets with the potential to block reactivation.

The experience of HIV reactive patients in rural Malawi - Part I

Directory of Open Access Journals (Sweden)

Y Sliep

2001-09-01

Full Text Available Malawi has a population of 9 million people with AIDS the leading cause of death in the 20 - 40 age group. The HIV positive prevalence rate, estimated at 23% in urban areas and 8% in rural areas, is one of the highest in the world (AIDSEC, 1994:1. Evaluation of counselling practices showed poor results with counsellors feeling ineffective and inadequate. Patients are mostly tested on medical indication but patients who do not see the benefit of knowing their HIV status increasingly refuse testing. The counselling practise as it is known in the Western world is a foreign concept for patients living in rural Malawi. The high stigma of AIDS complicates support of the patients. The goal of the research study was to describe a model of counselling that would meet the needs of an AIDS patient in rural community in Malawi. A qualitative research design that was explorative, descriptive and contextually specific to rural Malawi was used for the study. In order to describe a counselling model it was important to understand the illness experience of HIV reactive patients. The patients are seen in group context congruent with the African culture and therefore the experience of the primary care giver of AIDS patients is explored as the other major factor in the phenomenon examined. One phase of the research is described in this article namely exploring and describing the experience of the HIV reactive patient in rural Malawi. Results show that patients are in an advanced stage of AIDS when they are diagnosed and complain of weakness and an inability to do work, including an inability to do their daily chores. This causes a feeling of desperateness that is worsened by the perception that support systems are inadequate. Support systems are mostly identified as parents, partners and siblings to assist mainly with the physical care and financial support. Despite the fact that the family is very important to patients there is a reluctance to acknowledge their

Transient ischemic attack presenting in an elderly patient with transient ophthalmic manifestations

Directory of Open Access Journals (Sweden)

Sparshi Jain

2016-01-01

Full Text Available Transient ischemic attack (TIA is a transient neurological deficit of cerebrovascular origin without infarction which may last only for a short period and can have varying presentations. We report a case of 58-year-old male with presenting features of sudden onset transient vertical diplopia and transient rotatory nystagmus which self-resolved within 12 h. Patient had no history of any systemic illness. On investigating, hematological investigations and neuroimaging could not explain these sudden and transient findings. A TIA could possibly explain these sudden and transient ocular findings in our patient. This case report aims to highlight the importance of TIA for ophthalmologists. We must not ignore these findings as these could be warning signs of an impending stroke which may or may not be detected on neuroimaging. Thus, early recognition, primary prevention strategies, and timely intervention are needed.

Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

International Nuclear Information System (INIS)

Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

2004-01-01

Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

Deformation modeling and the strain transient dip test

International Nuclear Information System (INIS)

Jones, W.B.; Rohde, R.W.; Swearengen, J.C.

1980-01-01

Recent efforts in material deformation modeling reveal a trend toward unifying creep and plasticity with a single rate-dependent formulation. While such models can describe actual material deformation, most require a number of different experiments to generate model parameter information. Recently, however, a new model has been proposed in which most of the requisite constants may be found by examining creep transients brought about through abrupt changes in creep stress (strain transient dip test). The critical measurement in this test is the absence of a resolvable creep rate after a stress drop. As a consequence, the result is extraordinarily sensitive to strain resolution as well as machine mechanical response. This paper presents the design of a machine in which these spurious effects have been minimized and discusses the nature of the strain transient dip test using the example of aluminum. It is concluded that the strain transient dip test is not useful as the primary test for verifying any micromechanical model of deformation. Nevertheless, if a model can be developed which is verifiable by other experimentts, data from a dip test machine may be used to generate model parameters

SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR

International Nuclear Information System (INIS)

Halsall, M.J.; Course, A.F.; Sidell, J.

1979-09-01

SPLOSH is a time dependent, one dimensional, finite difference (in time and space) coupled neutron kinetics and thermal hydraulics code for studying pressurised faults and control transients in water reactor systems. An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the reactor core. A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface. The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly. Circuit features modelled include pumps, spray cooling and the SGHWR steam drum. Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified. (UK)

Mitigation method of thermal transient stress by thermalhydraulic-structure total analysis

International Nuclear Information System (INIS)

Kasahara, Naoto; Jinbo, Masakazu; Hosogai, Hiromi

2003-01-01

This study proposes a rational evaluation and mitigation method of thermal transient loads in fast reactor components by utilizing relationships among plant system parameters and stresses induced by thermal transients of plants. A thermalhydraulic-structure total analysis procedure helps us to grasp relationship among system parameters and thermal stresses. Furthermore, it enables mitigation of thermal transient loads by adjusting system parameters. In order to overcome huge computations, a thermalhydraulic-structure total analysis code and the Design of Experiments methodology are utilized. The efficiency of the proposed mitigation method is validated through thermal stress evaluation of an intermediate heat exchanger in Japanese demonstration fast reactor. (author)

Staphyloxanthin photobleaching sensitizes methicillin-resistant Staphylococcus aureus to reactive oxygen species attack

Science.gov (United States)

Dong, Pu-Ting; Mohammad, Haroon; Hui, Jie; Wang, Xiaoyu; Li, Junjie; Liang, Lijia; Seleem, Mohamed N.; Cheng, Ji-Xin

2018-02-01

Given that the dearth of new antibiotic development loads an existential burden on successful infectious disease therapy, health organizations are calling for alternative approaches to combat methicillin-resistant Staphylococcus aureus (MRSA) infections. Here, we report a drug-free photonic approach to eliminate MRSA through photobleaching of staphyloxanthin, an indispensable membrane-bound antioxidant of S. aureus. The photobleaching process, uncovered through a transient absorption imaging study and quantitated by absorption spectroscopy and mass spectrometry, decomposes staphyloxanthin, and sensitizes MRSA to reactive oxygen species attack. Consequently, staphyloxanthin bleaching by low-level blue light eradicates MRSA synergistically with external or internal reactive oxygen species. The effectiveness of this synergistic therapy is validated in MRSA culture, MRSAinfected macrophage cells. Collectively, these findings highlight broad applications of staphyloxanthin photobleaching for treatment of MRSA infections.

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.

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)

Fast thermal transients on valve

International Nuclear Information System (INIS)

Ferjancic, M.; Stok, B.; Halilovic, M.; Koc, P.; Mole, N.; Otrin, Z.; Kotar, A.

2007-01-01

One of the regulatory body methods to supervise nuclear safety of a nuclear power plant is a review of plant modifications and evaluation of their impact on plant operating experience. The Slovenian Nuclear Safety Administration (SNSA) licensed in April 2003 the use of leak-before-break (LBB) methodology in the Krsko NPP for the primary loop including surge line and connecting pipelines with minimal diameter of 6 inch. The SNSA decision based also on fracture mechanics analyses that include direct pipe failure mechanisms such as water hammer, creep damage, erosion and corrosion, fatigue and environmental conditions over the entire life of the plant. The evaluation of the operating transients pointed out, that presumed loadings, used for the LBB analysis, did not incorporate all the fast thermal transients data. For that purpose the SNSA requested Faculty of Mechanical Engineering (FS) in Ljubljana to perform additional analyses. The results of the analysis shall confirm the validity of the LBB analysis. (author)

Neighborhood disadvantage and adolescent stress reactivity

Directory of Open Access Journals (Sweden)

Daniel A. Hackman

2012-10-01

Full Text Available Lower socioeconomic status (SES is associated with higher levels of life stress, which in turn affect stress physiology. SES is related to basal cortisol and diurnal change, but it is not clear if SES is associated with cortisol reactivity to stress. To address this question, we examined the relationship between two indices of SES, parental education and concentrated neighborhood disadvantage, and the cortisol reactivity of African-American adolescents to a modified version of the Trier Social Stress Test. We found that concentrated disadvantage was associated with cortisol reactivity and this relationship was moderated by gender, such that higher concentrated disadvantage predicted higher cortisol reactivity and steeper recovery in boys but not in girls. Parental education, alone or as moderated by gender, did not predict reactivity or recovery, while neither education nor concentrated disadvantage predicted estimates of baseline cortisol. This finding is consistent with animal literature showing differential vulnerability, by gender, to the effects of adverse early experience on stress regulation and the differential effects of neighborhood disadvantage in adolescent males and females. This suggests that the mechanisms underlying SES differences in brain development and particularly reactivity to environmental stressors may vary across genders.

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

Multi-dose Romidepsin Reactivates Replication Competent SIV in Post-antiretroviral Rhesus Macaque Controllers.

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Benjamin B Policicchio

2016-09-01

Full Text Available Viruses that persist despite seemingly effective antiretroviral treatment (ART and can reinitiate infection if treatment is stopped preclude definitive treatment of HIV-1 infected individuals, requiring lifelong ART. Among strategies proposed for targeting these viral reservoirs, the premise of the "shock and kill" strategy is to induce expression of latent proviruses [for example with histone deacetylase inhibitors (HDACis] resulting in elimination of the affected cells through viral cytolysis or immune clearance mechanisms. Yet, ex vivo studies reported that HDACis have variable efficacy for reactivating latent proviruses, and hinder immune functions. We developed a nonhuman primate model of post-treatment control of SIV through early and prolonged administration of ART and performed in vivo reactivation experiments in controller RMs, evaluating the ability of the HDACi romidepsin (RMD to reactivate SIV and the impact of RMD treatment on SIV-specific T cell responses. Ten RMs were IV-infected with a SIVsmmFTq transmitted-founder infectious molecular clone. Four RMs received conventional ART for >9 months, starting from 65 days post-infection. SIVsmmFTq plasma viremia was robustly controlled to <10 SIV RNA copies/mL with ART, without viral blips. At ART cessation, initial rebound viremia to ~106 copies/mL was followed by a decline to < 10 copies/mL, suggesting effective immune control. Three post-treatment controller RMs received three doses of RMD every 35-50 days, followed by in vivo experimental depletion of CD8+ cells using monoclonal antibody M-T807R1. RMD was well-tolerated and resulted in a rapid and massive surge in T cell activation, as well as significant virus rebounds (~104 copies/ml peaking at 5-12 days post-treatment. CD8+ cell depletion resulted in a more robust viral rebound (107 copies/ml that was controlled upon CD8+ T cell recovery. Our results show that RMD can reactivate SIV in vivo in the setting of post-ART viral control

Transient drainage summary report

International Nuclear Information System (INIS)

1996-09-01

This report summarizes the history of transient drainage issues on the Uranium Mill Tailings Remedial Action (UMTRA) Project. It defines and describes the UMTRA Project disposal cell transient drainage process and chronicles UMTRA Project treatment of the transient drainage phenomenon. Section 4.0 includes a conceptual cross section of each UMTRA Project disposal site and summarizes design and construction information, the ground water protection strategy, and the potential for transient drainage

TRANSIENT ELECTRONICS CATEGORIZATION

Science.gov (United States)

2017-08-24

AFRL-RY-WP-TR-2017-0169 TRANSIENT ELECTRONICS CATEGORIZATION Dr. Burhan Bayraktaroglu Devices for Sensing Branch Aerospace Components & Subsystems...SUBTITLE TRANSIENT ELECTRONICS CATEGORIZATION 5a. CONTRACT NUMBER In-house 5b. GRANT NUMBER N/A 5c. PROGRAM ELEMENT NUMBER N/A 6. AUTHOR(S) Dr. Burhan...88ABW-2017-3747, Clearance Date 31 July 2017. Paper contains color. 14. ABSTRACT Transient electronics is an emerging technology area that lacks proper

Automated Detection of Short Optical Transients of Astrophysical Origin in Real Time

Directory of Open Access Journals (Sweden)

Marcin Sokołowski

2010-01-01

Full Text Available The detection of short optical transients of astrophysical origin in real time is an important task for existing robotic telescopes. The faster a new optical transient is detected, the earlier follow-up observations can be started. The sooner the object is identified, the more data can be collected before the source fades away, particularly in the most interesting early period of the transient. In this the real-time pipeline designed for identification of optical flashes with the “Pi of the Sky” project will be presented in detail together with solutions used by other experiments.

How Life Experience Shapes Cognitive Control Strategies: The Case of Air Traffic Control Training.

Directory of Open Access Journals (Sweden)

Sandra Arbula

Full Text Available Although human flexible behavior relies on cognitive control, it would be implausible to assume that there is only one, general mode of cognitive control strategy adopted by all individuals. For instance, different reliance on proactive versus reactive control strategies could explain inter-individual variability. In particular, specific life experiences, like a highly demanding training for future Air Traffic Controllers (ATCs, could modulate cognitive control functions. A group of ATC trainees and a matched group of university students were tested longitudinally on task-switching and Stroop paradigms that allowed us to measure indices of cognitive control. The results showed that the ATCs, with respect to the control group, had substantially smaller mixing costs during long cue-target intervals (CTI and a reduced Stroop interference effect. However, this advantage was present also prior to the training phase. Being more capable in managing multiple task sets and less distracted by interfering events suggests a more efficient selection and maintenance of task relevant information as an inherent characteristic of the ATC group, associated with proactive control. Critically, the training that the ATCs underwent improved their accuracy in general and reduced response time switching costs during short CTIs only. These results indicate a training-induced change in reactive control, which is described as a transient process in charge of stimulus-driven task detection and resolution. This experience-based enhancement of reactive control strategy denotes how cognitive control and executive functions in general can be shaped by real-life training and underlines the importance of experience in explaining inter-individual variability in cognitive functioning.

How Life Experience Shapes Cognitive Control Strategies: The Case of Air Traffic Control Training.

Science.gov (United States)

Arbula, Sandra; Capizzi, Mariagrazia; Lombardo, Nicoletta; Vallesi, Antonino

2016-01-01

Although human flexible behavior relies on cognitive control, it would be implausible to assume that there is only one, general mode of cognitive control strategy adopted by all individuals. For instance, different reliance on proactive versus reactive control strategies could explain inter-individual variability. In particular, specific life experiences, like a highly demanding training for future Air Traffic Controllers (ATCs), could modulate cognitive control functions. A group of ATC trainees and a matched group of university students were tested longitudinally on task-switching and Stroop paradigms that allowed us to measure indices of cognitive control. The results showed that the ATCs, with respect to the control group, had substantially smaller mixing costs during long cue-target intervals (CTI) and a reduced Stroop interference effect. However, this advantage was present also prior to the training phase. Being more capable in managing multiple task sets and less distracted by interfering events suggests a more efficient selection and maintenance of task relevant information as an inherent characteristic of the ATC group, associated with proactive control. Critically, the training that the ATCs underwent improved their accuracy in general and reduced response time switching costs during short CTIs only. These results indicate a training-induced change in reactive control, which is described as a transient process in charge of stimulus-driven task detection and resolution. This experience-based enhancement of reactive control strategy denotes how cognitive control and executive functions in general can be shaped by real-life training and underlines the importance of experience in explaining inter-individual variability in cognitive functioning.

Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

International Nuclear Information System (INIS)

Shin, Hyeong-Ki

1999-01-01

The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations

Diffusive–Dispersive and Reactive Fronts in Porous Media

DEFF Research Database (Denmark)

Haberer, Christina M.; Muniruzzaman, Muhammad; Grathwohl, Peter

2015-01-01

, across the unsaturated–saturated interface, under both conservative and reactive transport conditions. As reactive system we considered the abiotic oxidation of Fe2+ in the presence of O2. We studied the reaction kinetics in batch experiments and its coupling with diffusive and dispersive transport...... processes by means of one-dimensional columns and two-dimensional flow-through experiments, respectively. A noninvasive optode technique was used to track O2 transport into the initially anoxic porous medium at highly resolved spatial and temporal scales. The results show significant differences...

Awake reactivation predicts memory in humans

OpenAIRE

Staresina, Bernhard P.; Alink, Arjen; Kriegeskorte, Nikolaus; Henson, Richard N.

2013-01-01

How is new information converted into a memory trace? Here, we used functional neuroimaging to assess what happens to representations of new events after we first experience them. We found that a particular part of the medial temporal lobe, a brain region known to be critical for intact memory, spontaneously reactivates these events even when we are engaged in unrelated activities. Indeed, the extent to which such automatic reactivation occurs seems directly related to later memory performanc...

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

Transient and intermittent magnetic reconnections in TS-3/UTST merging startup experiments

International Nuclear Information System (INIS)

Ono, Y.; Imazawa, R.; Imanaka, H.; Hayamizu, T.; Inomoto, M.; Sato, M.; Kawamori, E.; Ejiri, A.; Takase, Y.; Asai, T.; Takahashi, T.

2007-01-01

The high-power reconnection heating has been developed in the TS-3 merging experiments, leading us to a new pulsed high-beta spherical tokamak (ST) formation. Two ST plasmas were produced inductively by two or four PF coils without using any central solenoid (CS) coil and were merged together for MW-GW reconnection heating. The magnetic reconnection transformed the magnetic energy of reconnecting magnetic field through the outflow kinetic energy finally to the ion thermal energy, increasing the plasma beta of ST up to 0.5. A new finding is that ejection of current sheet (or plasmoid) causes high-speed merging/ reconnection as well as high-power heating. In the high-q ST merging, the sheet resistivity was almost classical due to the sheet thickness much longer than ion gyroradius. Large inflow flux and low current-sheet dissipation resulted in flux pileup followed by rapid growth of the current sheet. When the flux pileup exceeded a critical limit, the sheet was ejected mechanically from the squeezed X-point area. The reconnection (outflow) speed was slow during the flux pileup and was fast during the ejection, indicating that intermittent reconnection similar to the solar flare increased the averaged reconnection speed. These transient effects enable us to have the fast reconnection as well as the high-power reconnection heating, even if the merging high-q tokamaks have low current-sheet resistivity. (author)

Volume 3. Base technology FSAR support document: prefailure transient behavior and failure threshold. Status report, January 1975

International Nuclear Information System (INIS)

Baars, R.E.; Culley, G.E.; Davis, R.T.; Henderson, R.G.; Scott, J.H.

1975-11-01

The FFTF fuel pin design and fabrication data, irradiation histories, and postirradiation examination results are summarized for the HEDL/TREAT transient test experimental programs. For each experiment, the data include: (a) fuel pin fabrication parameters, (b) steady-state irradiation history, (c) transient test design conditions, (d) transient test data, and (e) postirradiation examination results

A novel transient rotor current control scheme of a doubly-fed induction generator equipped with superconducting magnetic energy storage for voltage and frequency support

Science.gov (United States)

Shen, Yang-Wu; Ke, De-Ping; Sun, Yuan-Zhang; Daniel, Kirschen; Wang, Yi-Shen; Hu, Yuan-Chao

2015-07-01

A novel transient rotor current control scheme is proposed in this paper for a doubly-fed induction generator (DFIG) equipped with a superconducting magnetic energy storage (SMES) device to enhance its transient voltage and frequency support capacity during grid faults. The SMES connected to the DC-link capacitor of the DFIG is controlled to regulate the transient dc-link voltage so that the whole capacity of the grid side converter (GSC) is dedicated to injecting reactive power to the grid for the transient voltage support. However, the rotor-side converter (RSC) has different control tasks for different periods of the grid fault. Firstly, for Period I, the RSC injects the demagnetizing current to ensure the controllability of the rotor voltage. Then, since the dc stator flux degenerates rapidly in Period II, the required demagnetizing current is low in Period II and the RSC uses the spare capacity to additionally generate the reactive (priority) and active current so that the transient voltage capability is corroborated and the DFIG also positively responds to the system frequency dynamic at the earliest time. Finally, a small amount of demagnetizing current is provided after the fault clearance. Most of the RSC capacity is used to inject the active current to further support the frequency recovery of the system. Simulations are carried out on a simple power system with a wind farm. Comparisons with other commonly used control methods are performed to validate the proposed control method. Project supported by the National Natural Science Foundation of China (Grant No. 51307124) and the Major Program of the National Natural Science Foundation of China (Grant No. 51190105).

Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod

Energy Technology Data Exchange (ETDEWEB)

Pazirandeh, Ali [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Science and Research Branch; Hooshyar Mobaraki, Almas

2017-07-15

The safe operation of a reactor is based on feedback models. In this paper we attempted to discuss the influence of a non-uniform radial temperature distribution on the fuel rod temperature coefficient of reactivity. The paper demonstrates that the neutron properties of a reactor core is based on effective temperature of the fuel to obtain the correct fuel temperature feedback. The value of volume-averaged temperature being used in the calculations of neutron physics with feedbacks would result in underestimating the probable event. In the calculation it is necessary to use the effective temperature of the fuel in order to provide correct accounting of the fuel temperature feedback. Fuel temperature changes in different zones of the core and consequently reactivity coefficient change are an important parameter for analysis of transient conditions. The restricting factor that compensates the inserted reactivity is the temperature reactivity coefficient and effective delayed neutron fraction.

Steam-chemical reactivity for irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Anderl, R.A.; McCarthy, K.A.; Oates, M.A.; Petti, D.A.; Pawelko, R.J.; Smolik, G.R. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

1998-01-01

This paper reports the results of an experimental investigation to determine the influence of neutron irradiation effects and annealing on the chemical reactivity of beryllium exposed to steam. The work entailed measurements of the H{sub 2} generation rates for unirradiated and irradiated Be and for irradiated Be that had been previously annealed at different temperatures ranging from 450degC to 1200degC. H{sub 2} generation rates were similar for irradiated and unirradiated Be in steam-chemical reactivity experiments at temperatures between 450degC and 600degC. For irradiated Be exposed to steam at 700degC, the chemical reactivity accelerated rapidly and the specimen experienced a temperature excursion. Enhanced chemical reactivity at temperatures between 400degC and 600degC was observed for irradiated Be annealed at temperatures of 700degC and higher. This reactivity enhancement could be accounted for by the increased specific surface area resulting from development of a surface-connected porosity in the irradiated-annealed Be. (author)

Boron dilution transients during natural circulation flow in PWR-Experiments and CFD simulations

Energy Technology Data Exchange (ETDEWEB)

Hoehne, Thomas [Forschungszentrum Dresden-Rossendorf (FZD)-Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)], E-mail: T.Hoehne@fzd.de; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf (FZD)-Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)

2008-08-15

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loops 1:5 scaled Rossendorf coolant mixing model (ROCOM) mixing test facility. In particular thermal hydraulics analyses have shown, that weakly borated condensate can accumulate in the pump loop seal of those loops, which do not receive a safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side, the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.

Reactivation of a cryptobiotic stream ecosystem in the McMurdo Dry Valleys, Antarctica: A long-term geomorphological experiment

Science.gov (United States)

McKnight, Diane M.; Tate, C.M.; Andrews, E.D.; Niyogi, D.K.; Cozzetto, K.; Welch, K.; Lyons, W.B.; Capone, D.G.

2007-01-01

The McMurdo Dry Valleys of Antarctica contain many glacial meltwater streams that flow for 6 to 12??weeks during the austral summer and link the glaciers to the lakes on the valley floors. Dry valley streams gain solutes longitudinally through weathering reactions and microbial processes occurring in the hyporheic zone. Some streams have thriving cyanobacterial mats. In streams with regular summer flow, the mats are freeze-dried through the winter and begin photosynthesizing with the onset of flow. To evaluate the longer term persistence of cyanobacterial mats, we diverted flow to an abandoned channel, which had not received substantial flow for approximately two decades. Monitoring of specific conductance showed that for the first 3??years after the diversion, the solute concentrations were greater in the reactivated channel than in most other dry valley streams. We observed that cyanobacterial mats became abundant in the reactivated channel within a week, indicating that the mats had been preserved in a cryptobiotic state in the channel. Over the next several years, these mats had high rates of productivity and nitrogen fixation compared to mats from other streams. Experiments in which mats from the reactivated channel and another stream were incubated in water from both of the streams indicated that the greater solute concentrations in the reactivated channel stimulated net primary productivity of mats from both streams. These stream-scale experimental results indicate that the cryptobiotic preservation of cyanobacterial mats in abandoned channels in the dry valleys allows for rapid response of these stream ecosystems to climatic and geomorphological change, similar to other arid zone stream ecosystems. ?? 2006 Elsevier B.V. All rights reserved.

Investigation of transient melting of tungsten by ELMs in ASDEX Upgrade

International Nuclear Information System (INIS)

Krieger, K; Sieglin, B; Balden, M; De Marne, P; Nille, D; Rohde, V; Faitsch, M; Giannone, L; Herrmann, A; Coenen, J W; Göths, B; Laggner, F; Matthews, G F; Dejarnac, R; Horacek, J; Komm, M; Pitts, R A; Ratynskaia, S; Thoren, E; Tolias, P

2017-01-01

Repetitive melting of tungsten by power transients originating from edge localized modes (ELMs) has been studied in the tokamak experiment ASDEX Upgrade. Tungsten samples were exposed to H-mode discharges at the outer divertor target plate using the Divertor Manipulator II system. The exposed sample was designed with an elevated sloped surface inclined against the incident magnetic field to increase the projected parallel power flux to a level were transient melting by ELMs would occur. Sample exposure was controlled by moving the outer strike point to the sample location. As extension to previous melt studies in the new experiment both the current flow from the sample to vessel potential and the local surface temperature were measured with sufficient time resolution to resolve individual ELMs. The experiment provided for the first time a direct link of current flow and surface temperature during transient ELM events. This allows to further constrain the MEMOS melt motion code predictions and to improve the validation of its underlying model assumptions. Post exposure ex situ analysis of the retrieved samples confirms the decreased melt motion observed at shallower magnetic field line to surface angles compared to that at leading edges exposed to the parallel power flux. (paper)

Measuring vascular reactivity with resting-state blood oxygenation level-dependent (BOLD) signal fluctuations: A potential alternative to the breath-holding challenge?

Science.gov (United States)

Jahanian, Hesamoddin; Christen, Thomas; Moseley, Michael E; Pajewski, Nicholas M; Wright, Clinton B; Tamura, Manjula K; Zaharchuk, Greg

2017-07-01

Measurement of the ability of blood vessels to dilate and constrict, known as vascular reactivity, is often performed with breath-holding tasks that transiently raise arterial blood carbon dioxide (P a CO 2 ) levels. However, following the proper commands for a breath-holding experiment may be difficult or impossible for many patients. In this study, we evaluated two approaches for obtaining vascular reactivity information using blood oxygenation level-dependent signal fluctuations obtained from resting-state functional magnetic resonance imaging data: physiological fluctuation regression and coefficient of variation of the resting-state functional magnetic resonance imaging signal. We studied a cohort of 28 older adults (69 ± 7 years) and found that six of them (21%) could not perform the breath-holding protocol, based on an objective comparison with an idealized respiratory waveform. In the subjects that could comply, we found a strong linear correlation between data extracted from spontaneous resting-state functional magnetic resonance imaging signal fluctuations and the blood oxygenation level-dependent percentage signal change during breath-holding challenge ( R 2  = 0.57 and 0.61 for resting-state physiological fluctuation regression and resting-state coefficient of variation methods, respectively). This technique may eliminate the need for subject cooperation, thus allowing the evaluation of vascular reactivity in a wider range of clinical and research conditions in which it may otherwise be impractical.

Intra-individual variability in cerebrovascular and respiratory chemosensitivity: Can we characterize a chemoreflex "reactivity profile"?

Science.gov (United States)

Borle, Kennedy J; Pfoh, Jamie R; Boulet, Lindsey M; Abrosimova, Maria; Tymko, Michael M; Skow, Rachel J; Varner, Amy; Day, Trevor A

2017-08-01

Intra-individual variability in the magnitude of human cerebrovascular and respiratory chemoreflex responses is largely unexplored. By comparing response magnitudes of cerebrovascular CO 2 reactivity (CVR; middle and posterior cerebral arteries; MCA, PCA), central (CCR; CO 2 ) and peripheral respiratory chemoreflexes (PCR; CO 2 and O 2 ), we tested the hypothesis that a within-individual reactivity magnitude profile could be characterized. The magnitudes of CVR and CCR were tested with hyperoxic rebreathing and PCR magnitudes were tested through transient respiratory tests (TT-CO 2 , hypercapnia; TT-N 2 , hypoxia). No significant intra-individual relationships were found between CCR vs. CVR (MCA and PCA), CCR vs. PCR (TT-N 2 or TT-CO 2 ) (r0.3) response magnitudes. Statistically significant relationships were found between MCA vs. PCA reactivity (r=0.45, Pvariability that exists in human cerebrovascular and respiratory chemoreflexes. Copyright © 2017 Elsevier B.V. All rights reserved.

PSH Transient Simulation Modeling

Energy Technology Data Exchange (ETDEWEB)

Muljadi, Eduard [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-12-21

PSH Transient Simulation Modeling presentation from the WPTO FY14 - FY16 Peer Review. Transient effects are an important consideration when designing a PSH system, yet numerical techniques for hydraulic transient analysis still need improvements for adjustable-speed (AS) reversible pump-turbine applications.

Improvement of the dynamic response of the ITER Reactive Power Compensation system

International Nuclear Information System (INIS)

Finotti, Claudio; Gaio, Elena; Song, Inho; Tao, Jun; Benfatto, Ivone

2015-01-01

Highlights: • The slow response reasons of the classic ITER Reactive Power Compensation (RPC) control are explained. • The dynamic behaviors of the ac/dc converter and of the RPC are characterized. • New control concept to speed up the RPC response is developed. • Good performance of the new RPC control is verified even during fast transient conditions. - Abstract: The ITER ac/dc conversion system can absorb a total active and reactive power up to 500 MW and 950 Mvar, respectively. The Reactive Power Compensation (RPC) system is rated for a nominal power of 750 Mvar necessary to comply with the allowable reactive power limit value from the grid of 200 Mvar. This system is currently under construction and is based on Static Var Compensation technology with Thyristor Controlled Reactor (TCR) and Tuned Filters. The RPC has to minimize the demand of reactive power from the grid; its control is based on a feed-forward method, where the corrective input is the measurement of the reactive power consumption of the ac/dc converters, derived from the 50 Hz component of the Fast Fourier Transform (FFT) of the three-phase voltages and currents. The delay introduced by the FFT calculation and the slow response of the TCR could make the response speed of the RPC not sufficient to face fast variations of the reactive power demand and therefore in this paper a new controller of the RPC able to overcome this shortcoming is proposed and evaluated. It is based on the calculation of the predicted consumption of the reactive power by using the voltage reference signals coming from the Plasma Control System and the measurements of the dc current of the ac/dc converters and of the 66 kV busbar voltage, and on the speed up of the RPC control by introducing a lead–lag transfer function.

Improvement of the dynamic response of the ITER Reactive Power Compensation system

Energy Technology Data Exchange (ETDEWEB)

Finotti, Claudio, E-mail: claudio.finotti@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy); Gaio, Elena [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy); Song, Inho; Tao, Jun; Benfatto, Ivone [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2015-10-15

Highlights: • The slow response reasons of the classic ITER Reactive Power Compensation (RPC) control are explained. • The dynamic behaviors of the ac/dc converter and of the RPC are characterized. • New control concept to speed up the RPC response is developed. • Good performance of the new RPC control is verified even during fast transient conditions. - Abstract: The ITER ac/dc conversion system can absorb a total active and reactive power up to 500 MW and 950 Mvar, respectively. The Reactive Power Compensation (RPC) system is rated for a nominal power of 750 Mvar necessary to comply with the allowable reactive power limit value from the grid of 200 Mvar. This system is currently under construction and is based on Static Var Compensation technology with Thyristor Controlled Reactor (TCR) and Tuned Filters. The RPC has to minimize the demand of reactive power from the grid; its control is based on a feed-forward method, where the corrective input is the measurement of the reactive power consumption of the ac/dc converters, derived from the 50 Hz component of the Fast Fourier Transform (FFT) of the three-phase voltages and currents. The delay introduced by the FFT calculation and the slow response of the TCR could make the response speed of the RPC not sufficient to face fast variations of the reactive power demand and therefore in this paper a new controller of the RPC able to overcome this shortcoming is proposed and evaluated. It is based on the calculation of the predicted consumption of the reactive power by using the voltage reference signals coming from the Plasma Control System and the measurements of the dc current of the ac/dc converters and of the 66 kV busbar voltage, and on the speed up of the RPC control by introducing a lead–lag transfer function.

Reactive model for developing applications using Vert.x toolkit

OpenAIRE

Ožbot, Žan

2017-01-01

Web and mobile applications consist of real-time events of different kinds in order to ensure the best possible user experience. To develop such applications, proper tools are needed and reactive programming is one of the possible solutions. Due to its many advantages, reactive programming is becoming an increasing reason to abandon standard object-oriented approach. Therefore, in this thesis we first describe the concepts of reactive programming and compare it to object-oriented programming....

Using Novel Laboratory Incubations and Field Experiments to Identify the Source and Fate of Reactive Organic Carbon in an Arsenic-contaminated Aquifer System

Science.gov (United States)

Stahl, M.; Tarek, M. H.; Badruzzaman, B.; Harvey, C. F.

2017-12-01

Characterizing the sources and fate of organic matter (OM) within aquifer systems is key to our understanding of both the broader global carbon cycle as well as the quality of our groundwater resources. The linkage between the subsurface carbon cycle and groundwater quality is perhaps nowhere more apparent than in the aquifer systems of South and Southeast Asia, where the contamination of groundwater with geogenic arsenic (As) is widespread and threatens the health of millions of individuals. OM fuels the biogeochemical processes driving As mobilization within these aquifers, however the source (i.e., modern surface-derived or aged sedimentary OM) of the reactive OM is widely debated. To characterize the sources of OM driving aquifer redox processes we tracked DIC and DOC concentrations and isotopes (stable and radiocarbon) along groundwater flow-paths and beneath an instrumented study pond at a field site in Bangladesh. We also conducted a set of novel groundwater incubation experiments, where we carbon-dated the DOC at the start and end of a experiment in order to determine the age of the OM that was mineralized. Our carbon/isotope balance reveals that aquifer recharge introduces a large quantity of young (i.e. near modern) OM that is efficiently mineralized within the upper few meters of the aquifer, effectively limiting this pool of reactive surface-sourced OM from being transported deeper into the aquifer where significant As mobilization takes place. The OM mineralized past the upper few meters is an aged, sedimentary source. Consistent with our field data, our incubation experiments show that past the upper few meters of the aquifer the reactive DOC is significantly older than the bulk DOC and has an age consistent with sedimentary OM. Combining our novel set of incubation experiments and a carbon/isotope balance along groundwater flow-paths and beneath our study pond we have identified the sources of reactive OM across different aquifer depths in a

Inverse Transient Analysis for Classification of Wall Thickness Variations in Pipelines

Directory of Open Access Journals (Sweden)

Jeffrey Tuck

2013-12-01

Full Text Available Analysis of transient fluid pressure signals has been investigated as an alternative method of fault detection in pipeline systems and has shown promise in both laboratory and field trials. The advantage of the method is that it can potentially provide a fast and cost effective means of locating faults such as leaks, blockages and pipeline wall degradation within a pipeline while the system remains fully operational. The only requirement is that high speed pressure sensors are placed in contact with the fluid. Further development of the method requires detailed numerical models and enhanced understanding of transient flow within a pipeline where variations in pipeline condition and geometry occur. One such variation commonly encountered is the degradation or thinning of pipe walls, which can increase the susceptible of a pipeline to leak development. This paper aims to improve transient-based fault detection methods by investigating how changes in pipe wall thickness will affect the transient behaviour of a system; this is done through the analysis of laboratory experiments. The laboratory experiments are carried out on a stainless steel pipeline of constant outside diameter, into which a pipe section of variable wall thickness is inserted. In order to detect the location and severity of these changes in wall conditions within the laboratory system an inverse transient analysis procedure is employed which considers independent variations in wavespeed and diameter. Inverse transient analyses are carried out using a genetic algorithm optimisation routine to match the response from a one-dimensional method of characteristics transient model to the experimental time domain pressure responses. The accuracy of the detection technique is evaluated and benefits associated with various simplifying assumptions and simulation run times are investigated. It is found that for the case investigated, changes in the wavespeed and nominal diameter of the

Inverse Transient Analysis for Classification of Wall Thickness Variations in Pipelines

Science.gov (United States)

Tuck, Jeffrey; Lee, Pedro

2013-01-01

Analysis of transient fluid pressure signals has been investigated as an alternative method of fault detection in pipeline systems and has shown promise in both laboratory and field trials. The advantage of the method is that it can potentially provide a fast and cost effective means of locating faults such as leaks, blockages and pipeline wall degradation within a pipeline while the system remains fully operational. The only requirement is that high speed pressure sensors are placed in contact with the fluid. Further development of the method requires detailed numerical models and enhanced understanding of transient flow within a pipeline where variations in pipeline condition and geometry occur. One such variation commonly encountered is the degradation or thinning of pipe walls, which can increase the susceptible of a pipeline to leak development. This paper aims to improve transient-based fault detection methods by investigating how changes in pipe wall thickness will affect the transient behaviour of a system; this is done through the analysis of laboratory experiments. The laboratory experiments are carried out on a stainless steel pipeline of constant outside diameter, into which a pipe section of variable wall thickness is inserted. In order to detect the location and severity of these changes in wall conditions within the laboratory system an inverse transient analysis procedure is employed which considers independent variations in wavespeed and diameter. Inverse transient analyses are carried out using a genetic algorithm optimisation routine to match the response from a one-dimensional method of characteristics transient model to the experimental time domain pressure responses. The accuracy of the detection technique is evaluated and benefits associated with various simplifying assumptions and simulation run times are investigated. It is found that for the case investigated, changes in the wavespeed and nominal diameter of the pipeline are both important

The effects of transient conditions on the onset of intermittent dryout during blowdown

Energy Technology Data Exchange (ETDEWEB)

Statham, B.A., E-mail: stathaba@mcmaster.ca; Novog, D.R., E-mail: novog@mcmaster.ca

2017-06-15

Highlights: • This papers presents the results of an experimental investigation of transient critical heat flux in high quality and intermediate pressure water. • In existing literature conclusions vary from those showing no effect of transient conditions to results which show 30–40% improvement in CHF. • Along with new CHF data points in the liquid film dominated flow regime, the authors provide a methodology for producing bias free estimates of CHF based on existing correlations. • With these bias free CHF estimates, comparisons are made between transient and steady-state CHF at comparable local conditions. • The work concludes that based on consistently collected and analyzed data that quasi-steady CHF experiments adequately predict transient CHF using the same local thermalhydraulic conditions. - Abstract: For a given set of conditions in a boiling system the point of liquid film dryout or departure from nucleate boiling corresponds to the change from convective or nucleate boiling to transition or film boiling. This change is associated with a rapid deterioration of the heat transfer coefficient and the heat flux at this transition is denoted the critical heat flux (CHF). Computer models used to predict station transients and CHF rely heavily on empirical correlations to predict the CHF. Liquid film CHF data are usually obtained using a quasi-steady method wherein the heat flux is incremented in small steps with each step being allowed to reach a new equilibrium until an abnormal temperature increase is detected on the experimental surfaces. In applying a correlation derived from steady-state experiments to transient analyses these codes implicitly assume that dryout will occur for the same local conditions during transients as during steady state conditions. There is some disagreement in literature as to the validity of this hypothesis. This paper provides new steady-state and transient experimental data for CHF in water at intermediate pressures

Neural mechanisms of reactivation-induced updating that enhance and distort memory

OpenAIRE

St. Jacques, Peggy L.; Olm, Christopher; Schacter, Daniel L.

2013-01-01

We remember a considerable number of personal experiences because we are frequently reminded of them, a process known as memory reactivation. Although memory reactivation helps to stabilize and update memories, reactivation may also introduce distortions if novel information becomes incorporated with memory. Here we used functional magnetic resonance imaging (fMRI) to investigate the neural mechanisms mediating reactivation-induced updating in memory for events experienced during a museum tou...

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.

Acetylene and carbon monoxide oxidation over a Pt/Rh/CeO2/γ-Al2O3 automotive exhaust gas catalyst: kinetic modelling of transient experiments

NARCIS (Netherlands)

Harmsen, J.M.A.; Hoebink, J.H.B.J.; Schouten, J.C.

2001-01-01

The transient kinetics of acetylene (C2H2) conversion by oxygen over a commercial Pt/Rh/CeO2/¿-Al2O3 three-way catalyst have been modelled. Experiments to validate the model were carried out in a fixed-bed reactor with two separate inlets, enabling alternate feeding of acetylene and oxygen.

Reactivity worth measurements on fast burst reactor Caliban - description and interpretation of integral experiments for the validation of nuclear data

Energy Technology Data Exchange (ETDEWEB)

Richard, B. [Commissariat a l' Energie Atomique et Aux Energies Alternatives CEA, DAM, VALDUC, F-21120 Is-sur-Tille (France)

2012-07-01

Reactivity perturbation experiments using various materials are being performed on the HEU fast core CALIBAN, an experimental device operated by the CEA VALDUC Criticality and Neutron Transport Research Laboratory. These experiments provide valuable information to contribute to the validation of nuclear data for the materials used in such measurements. This paper presents the results obtained in a first series of measurements performed with Au-197 samples. Experiments which have been conducted in order to improve the characterization of the core are also described and discussed. The experimental results have been compared to numerical calculation using both deterministic and Monte Carlo neutron transport codes with a simplified model of the reactor. This early work led to a methodology which will be applied to the future experiments which will concern other materials of interest. (authors)

Reactive flow modeling of small scale detonation failure experiments for a baseline non-ideal explosive

Energy Technology Data Exchange (ETDEWEB)

Kittell, David E.; Cummock, Nick R.; Son, Steven F. [School of Mechanical Engineering, Purdue University, West Lafayette, Indiana 47907 (United States)

2016-08-14

Small scale characterization experiments using only 1–5 g of a baseline ammonium nitrate plus fuel oil (ANFO) explosive are discussed and simulated using an ignition and growth reactive flow model. There exists a strong need for the small scale characterization of non-ideal explosives in order to adequately survey the wide parameter space in sample composition, density, and microstructure of these materials. However, it is largely unknown in the scientific community whether any useful or meaningful result may be obtained from detonation failure, and whether a minimum sample size or level of confinement exists for the experiments. In this work, it is shown that the parameters of an ignition and growth rate law may be calibrated using the small scale data, which is obtained from a 35 GHz microwave interferometer. Calibration is feasible when the samples are heavily confined and overdriven; this conclusion is supported with detailed simulation output, including pressure and reaction contours inside the ANFO samples. The resulting shock wave velocity is most likely a combined chemical-mechanical response, and simulations of these experiments require an accurate unreacted equation of state (EOS) in addition to the calibrated reaction rate. Other experiments are proposed to gain further insight into the detonation failure data, as well as to help discriminate between the role of the EOS and reaction rate in predicting the measured outcome.

Modeling of the transient mobility in disordered organic semiconductors

NARCIS (Netherlands)

Germs, W.C.; Van der Holst, J.M.M.; Van Mensfoort, S.L.M.; Bobbert, P.A.; Coehoorn, R.

2011-01-01

In non-steady-state experiments, the electrical response of devicesbased on disordered organic semiconductors often shows a large transient contribution due to relaxation of the out-of-equilibrium charge-carrier distribution. We have developed a model describing this process, based only on the

The behaviour of irradiated fuel under RIA transients: Interpretation of the CABRI experiments

International Nuclear Information System (INIS)

Papin, J.; Rigat, H.; Breton, J.P.; Schmitz, F.

1996-01-01

Paper presents the results of investigation of highly irradiated PWR fuel behaviour under fast power transients conducted in a sodium loop of CABRI reactor, as well as the results on development and validation of computer code SCANAIR. (author). 8 refs, 9 figs, 2 tabs

Reinstatement versus Reactivation Effects on Active Memory in Infants.

Science.gov (United States)

Adler, Scott A.; Rovee-Collier, Carolyn; Wilk, Amy

2000-01-01

Four experiments examined whether reinstatement and reactivation reminder paradigms affected memory performance of 102 three-month-olds. Results indicated that a single reinstatement protracted retention twice as long after training as a single reactivation. The novelty of the reminder stimulus also affected duration and specificity of memory in…

Reactivity Accidents in CAREM-25 Core with and Without Safety Systems Actuation

International Nuclear Information System (INIS)

Gimenez, Marcelo; Vertullo, Alicia; Schlamp, Miguel

2000-01-01

A reactivity accident in CAREM core can be provoked by different initiating events, a cold water injection in pressure vessel, a secondary side steam line breakage and a failure in the absorbing rods drive system.The present work analyses inadverted control rod withdraws transients.Maximum worth control rod (2.5 $) at normal velocity (1 cm/s) is adopted for the simulations (Reactivity ramp of 0.018 $/s).Different scenarios considering actuation of first shutdown system (FSS), second shutdown system (SSS) and selflimiting conditions were modeled.Results of the accident with actuation of FSS show that safety margins are well above critical values (DNBR and CPR).In the cases with failure of the FSS and success of SSS or selflimited, safety margins are below critical values, however, the SSS provides a reduction of elapsed time under advised margins

Transient bowing of core assemblies in advanced liquid metal fast reactors

International Nuclear Information System (INIS)

Kamal, S.A.; Orechwa, Y.

1986-01-01

Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

Transient analysis on the SMART-P anticipated transients without scram

International Nuclear Information System (INIS)

Yang, S. H.; Bae, K. H.; Kim, H. C.; Zee, S. Q.

2005-01-01

Anticipated transients without scram (ATWS) are anticipated operational occurrences accompanied by a failure of an automatic reactor trip when required. Although the occurrence probability of the ATWS events is considerably low, these events can result in unacceptable consequences, i.e. the pressurization of the reactor coolant system (RCS) up to an unacceptable range and a core-melting situation. Therefore, the regulatory body requests the installation of a protection system against the ATWS events. According to the request, a diverse protection system (DPS) is installed in the SMART-P (System-integrated Modular Advanced ReacTor-Pilot). This paper presents the results of the transient analysis performed to identify the performance of the SMART-P against the ATWS. In the analysis, the TASS/SMR (Transients And Setpoint Simulation/Small and Medium Reactor) code is applied to identify the thermal hydraulic response of the RCS during the transients

A transient single particle model under FCI conditions

Institute of Scientific and Technical Information of China (English)

LI Xiao-Yan; SHANG Zhi; XU Ji-Jun

2005-01-01

The paper is focused on the coupling effect between film boiling heat transfer and evaporation drag around a hot-particle in cold liquid. Based on the continuity, momentum and energy equations of the vapor film, a transient two-dimensional single particle model has been established. This paper contains a detailed description of HPMC (High-temperature Particle Moving in Coolant) model for studying some aspects of the premixing stage of fuel-coolant interactions (FCIs). The transient process of high-temperature particles moving in coolant can be simulated. Comparisons between the experiment results and the calculations using HPMC model demonstrate that HPMC model achieves a good agreement in predicting the time-varying characteristic of high-temperature spheres moving in coolant.

A revision of sensitivity analysis for small reactivity effects in ZPRs

International Nuclear Information System (INIS)

Ros, Paul; Blaise, Patrick; Gruel, Adrien; Leconte, Pierre

2017-01-01

Sensitivity analysis appears to be an important element for nuclear data improvement experiments. Indeed, it brings significant information on the contribution of the isotopes involved in the measurements performed in Zero Power Reactors (ZPRs), particularly oscillation measurements like in MINERVE, and its successor ZEPHYR (Zero power Experimental PHYsics Reactor), currently being designed at CEA. Oscillation measurements consist in oscillating a small sample made of separated isotopes (or irradiated fuels) in the core center. Then, two perturbations occur: a local one corresponding to the flux modification around the sample, and a global one which corresponds to the induced variation of reactivity. This variation of reactivity is either uncontrolled (open loop) or automatically compensated by an external pilot rod (closed loop) to keep the configuration in its critical state. Representativity studies are used in order to evaluate the pertinence of an experiment configuration versus a targeted application. For oscillation experiments, sensitivity of the reactivity effects to nuclear data is needed to obtain such coefficients. The Equivalent Generalized Perturbation Theory (EGPT) method, based on an approximation of the Generalized Perturbation. Theory, is currently applied in the ERANOS code for control rod insertions and other important variations of reactivity. However, such reactivity insertions induce consequent reactivity changes and variations of the flux, whereas oscillations induce maximal reactivity effects of 10 pcm (10 10 -5 Δk/k ) and consequently very local variations of the flux surrounding the sample. Therefore, such numerical methods are not necessarily adapted to the calculation of small reactivity effect sensitivities to nuclear data. The influence of peripheral isotopes (through their cross-sections) to central measurements is evaluated thanks to the deterministic EGPT method and the Monte-Carlo technique of correlated samples. Large

Performance of high burned PWR fuel during transient

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio

1992-01-01

In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

The role of grain boundary fission gases in high burn-up fuel under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Lemoine, F.; Papin, J.; Frizonnet, J.M.; Cazalis, B.; Rigat, H.

2002-01-01

In the frame of reactivity-initiated accidents (RIA) studies, the CABRI REP-Na programme is currently performed, focused on high burn-up UO 2 and MOX fuel behaviour. From 1993 to 1998, seven tests were performed with UO 2 fuel and three with MOX fuel. In all these tests, particular attention has been devoted to the role of fission gases in transient fuel behaviour and in clad loading mechanisms. From the analysis of experimental results, some basic phenomena were identified and a better understanding of the transient fission gas behaviour was obtained in relation to the fuel and clad thermo-mechanical evolution in RIA, but also to the initial state of the fuel before the transient. A high burn-up effect linked to the increasing part of grain boundary gases is clearly evidenced in the final gas release, which would also significantly contribute to the clad loading mechanisms. (authors)

OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Fourth Workshop (V100-CT4)

International Nuclear Information System (INIS)

2006-01-01

The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. Since the previous coupled code benchmarks indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and

Nuclear Fuel Behaviour during Reactivity Initiated Accidents. Workshop Proceedings

International Nuclear Information System (INIS)

2010-01-01

A reactivity initiated accident (RIA) is a nuclear reactor accident that involves an unwanted increase in fission rate and reactor power. The power increase may damage the reactor core. The main objective of the workshop was to review the current status of the experimental and analytical studies of the fuel behavior during the RIA transients in PWR and BWR reactors and the acceptance criteria for RIA in use and under consideration. The workshop was organized in an opening session and 5 technical sessions: 1) Recent experimental results and experimental techniques used; 2) Modelling and Data Interpretation; 3) Code Assessment; 4) RIA Core Analysis and 5) Revision and application of safety criteria

Biological stress reactivity as an index of the two polarities of the experience model.

Science.gov (United States)

Silva, Jaime R; Vivanco-Carlevari, Anastassia; Barrientos, Mauricio; Martínez, Claudio; Salazar, Luis A; Krause, Mariane

2017-10-01

The two-polarities model of personality argues that experience is organized around two axes: interpersonal relatedness and self-definition. Differential emphasis on one of these poles defines adaptive and pathological experiences, generating anaclitic or introjective tendencies. The anaclitic pattern, on one hand, has been conceptually related with an exaggerated emphasis on interpersonal relatedness. On the other hand, the introjective pattern has been connected to high levels of self-criticism. The aim of this study was to investigate the psychophysiological basis for this relationship. Specifically, we hypothesized that the anaclitic individual should have a higher biological reactivity to stress (BRS), measured by the cortisol concentration in saliva, in an interpersonal stress induction protocol (Trier Social Stress Test). Contrary to what was expected, the results indicated that introjective participants presented a higher BSR than the anaclitic group. Interestingly, in contrast to their higher BSR, the introjective group reported a diminished subjective stress in relation to the average. In the anaclitic group, a tendency that goes in the opposite direction was found. Theoretical implications of these findings were discussed. Copyright © 2017 Elsevier Ltd. All rights reserved.

Comparison of LIFE-4 and TEMECH code predictions with TREAT transient test data

International Nuclear Information System (INIS)

Gneiting, B.C.; Bard, F.E.; Hunter, C.W.

1984-09-01

Transient tests in the TREAT reactor were performed on FFTF Reference design mixed-oxide fuel pins, most of which had received prior steady-state irradiation in the EBR-II reactor. These transient test results provide a data base for calibration and verification of fuel performance codes and for evaluation of processes that affect pin damage during transient events. This paper presents a comparison of the LIFE-4 and TEMECH fuel pin thermal/mechanical analysis codes with the results from 20 HEDL TREAT experiments, ten of which resulted in pin failure. Both the LIFE-4 and TEMECH codes provided an adequate representation of the thermal and mechanical data from the TREAT experiments. Also, a criterion for 50% probability of pin failure was developed for each code using an average cumulative damage fraction value calculated for the pins that failed. Both codes employ the two major cladding loading mechanisms of differential thermal expansion and central cavity pressurization which were demonstrated by the test results. However, a detailed evaluation of the code predictions shows that the two code systems weigh the loading mechanism differently to reach the same end points of the TREAT transient results

Forced convection mixing transients in the MITR core tank

International Nuclear Information System (INIS)

Hu, Lin-Wen; Meyer, J.E.; Bernard, J.A.

1995-01-01

This paper reports the results of forced convection mixing transient experiments that were studied in the core tank of the 5-MW Massachusetts Institute of Technology (MIT) nuclear reactor as part of the studies being conducted to support a facility upgrade to 10 MW

Measurements of transient electron density distributions by femtosecond X-ray diffraction; Messungen transienter Elektronendichteverteilungen durch Femtosekunden-Roentgenbeugung

Energy Technology Data Exchange (ETDEWEB)

Freyer, Benjamin

2013-05-02

This thesis concerns measurements of transient charge density maps by femtosecond X-ray diffraction. Different X-ray diffraction methods will be considered, particularly with regard to their application in femtosecond X-ray diffraction. The rotation method is commonly used in stationary X-ray diffraction. In the work in hand an X-ray diffraction experiment is demonstrated, which combines the method with ultrafast X-ray pulses. This experiment is the first implementation which makes use of the rotation method to map transient intensities of a multitude of Bragg reflections. As a prototype material Bismuth is used, which previously was studied frequently by femtosecond X-ray diffraction by measuring Bragg reflections successively. The experimental results of the present work are compared with the literature data. In the second part a powder-diffraction experiment will be presented, which is used to study the dynamics of the electron-density distribution on ultrafast time scales. The experiment investigates a transition metal complex after photoexcitation of the metal to ligand charge transfer state. Besides expected results, i. e. the change of the bond length between the metal and the ligand and the transfer of electronic charge from the metal to the ligand, a strong contribution of the anion to the charge transfer was found. Furthermore, the charge transfer has predominantly a cooperative character. That is, the excitation of a single complex causes an alteration of the charge density of several neighboring units. The results show that more than 30 transition-metal complexes and 60 anions contribute to the charge transfer. This collective response is a consequence of the strong coulomb interactions of the densely packed ions.

Theory of strong-field attosecond transient absorption

International Nuclear Information System (INIS)

Wu, Mengxi; Chen, Shaohao; Camp, Seth; Schafer, Kenneth J; Gaarde, Mette B

2016-01-01

Attosecond transient absorption is one of the promising new techniques being developed to exploit the availability of sub-femtosecond extreme ultraviolet (XUV) pulses to study the dynamics of the electron on its natural time scale. The temporal resolution in a transient absorption setup comes from the control of the relative delay and coherence between pump and probe pulses, while the spectral resolution comes from the characteristic width of the features that are being probed. In this review we focus on transient absorption scenarios where an attosecond pulse of XUV radiation creates a broadband excitation that is subsequently probed by a few cycle infrared (IR) laser. Because the attosecond XUV pulses are locked to the IR field cycle, the exchange of energy in the laser–matter interaction can be studied with unprecedented precision. We focus on the transient absorption by helium atoms of XUV radiation around the first ionization threshold, where we can simultaneoulsy solve the time-dependent Schrödinger equation for the single atom response and the Maxwell wave equation for the collective response of the nonlinear medium. We use a time-domain method that allows us to treat on an equal footing all the different linear and nonlinear processes by which the medium can exchange energy with the fields. We present several simple models, based on a few-level system interacting with a strong IR field, to explain many of the novel features found in attosecond transient absorption spectrograms. These include the presence of light-induced states, which demonstrate the ability to probe the dressed states of the atom. We also present a time-domain interpretation of the resonant pulse propagation features that appear in absorption spectra in dense, macroscopic media. We close by reviewing several recent experimental results that can be explained in terms of the models we discuss. Our aim is to present a road map for understanding future attosecond transient absorption

Daily Emotional and Physical Reactivity to Stressors Among Widowed and Married Older Adults

Science.gov (United States)

2014-01-01

Objectives. Widowhood may result in declines in health and potentially stressful changes to daily routines. However, little research has examined how daily stressors contribute to physical and emotional well-being in widowhood. The objectives of the current study were to examine daily stressor exposure and reactivity in widowed versus married older adults. Method. Participants included all 100 widowed and 342 married adults aged 65 and older from the National Study of Daily Experiences, a daily diary study from the second wave of the Midlife in the United States. Daily stressors were measured using the Daily Inventory of Stressful Events; multilevel modeling assessed daily reactivity to stressors using daily negative affect (emotional reactivity) and daily physical symptoms (physical reactivity) as outcomes. Results. Married participants reported more stressors in general, and specifically more interpersonal stressors (e.g., arguments). Both married and widowed participants were reactive to daily stressors. Married participants were physically and emotionally reactive to interpersonal stressors. Widowed participants were more physically reactive to home-related stressors. Discussion. Attention to the types of daily stressors that widowed older adults experience in daily life and the potential physical effects of daily stressors during widowhood may help to alleviate some of the physical distress that widowed older adults may experience. PMID:23685921

Daily emotional and physical reactivity to stressors among widowed and married older adults.

Science.gov (United States)

Hahn, Elizabeth A; Cichy, Kelly E; Small, Brent J; Almeida, David M

2014-01-01

Widowhood may result in declines in health and potentially stressful changes to daily routines. However, little research has examined how daily stressors contribute to physical and emotional well-being in widowhood. The objectives of the current study were to examine daily stressor exposure and reactivity in widowed versus married older adults. Participants included all 100 widowed and 342 married adults aged 65 and older from the National Study of Daily Experiences, a daily diary study from the second wave of the Midlife in the United States. Daily stressors were measured using the Daily Inventory of Stressful Events; multilevel modeling assessed daily reactivity to stressors using daily negative affect (emotional reactivity) and daily physical symptoms (physical reactivity) as outcomes. Married participants reported more stressors in general, and specifically more interpersonal stressors (e.g., arguments). Both married and widowed participants were reactive to daily stressors. Married participants were physically and emotionally reactive to interpersonal stressors. Widowed participants were more physically reactive to home-related stressors. Attention to the types of daily stressors that widowed older adults experience in daily life and the potential physical effects of daily stressors during widowhood may help to alleviate some of the physical distress that widowed older adults may experience.

Simulation of LOFT anticipated-transient experiments L6-1, L6-2, and L6-3 using TRAC-PF1/MOD1

International Nuclear Information System (INIS)

Sahota, M.S.

1984-01-01

Anticipated-transient experiments L6-1, L6-2, and L6-3, performed at the Loss-of-fluid Test (LOFT) facility, are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1/MOD1). The results are used to assess TRAC-PF1/MOD1 trip and control capabilities, and predictions of thermal-hydraulic phenomena during slow transients. Test L6-1 simulated a loss-of-stream load in a large pressurized-water reactor (PWR), and was initiated by closing the main steam-flow control valve (MSFCV) at its maximum rate, which reduced the heat removal from the secondary-coolant system and increased the primary-coolant system pressure that initiated a reactor scram. Test L6-2 simulated a loss-of-primary coolant flow in a large PWR, and was initiated by tripping the power to the primary-coolant pumps (PCPs) allowing the pumps to coast down. The reduced primary-coolant flow caused a reactor scram. Test L6-3 simulated an excessive-load increase incident in a large PWR, and was initiated by opening the MSFCV at its maximum rate, which increased the heat removal from the secondary-coolant system and decreased the primary-coolant system pressure that initiated a reactor scram. The TRAC calculations accurately predict most test events. The test data and the calculated results for most parameters of interest also agree well

Development of advanced BWR fuel bundle with spectral shift rod (3) -transient analysis of ABWR core with SSR

International Nuclear Information System (INIS)

Ikegawa, T.; Chaki, M.; Ohga, Y.; Abe, M.

2010-01-01

The spectral shift rod (SSR) is a new type of water rod, utilized instead of the conventional water rod, in which a water level develops during core operation. The water level can be changed according to the fuel channel flow rate. In this study, ABWR plant performance with SSR fuel bundles under transient conditions has been evaluated using the TRACG code. The TRACG code, which can treat three-dimensional hydrodynamic calculations in a reactor pressure vessel, is well suited for evaluating the reactor transient performance with the SSR fuel bundles because it can calculate the water levels in the SSR at each channel grouping and therefore evaluate the core reactivity according to the water level changes in the SSR. 'Generator load rejection with total turbine bypass failure' and 'Recirculation flow control failure with increasing flow' were selected as cases which may increase the reactivity with the increasing water level in the SSR. It was found that the absolute value of the void reactivity coefficient in the SSR core was larger than that in the conventional water rod core because the core averaged void fraction in the SSR core, which has the vapor region above the water level in the SSR, was larger than that in the conventional water rod core. Therefore, AMCPR for the SSR core was a little larger than that for the conventional water rod core; however, the difference was smaller than 0.02 because the inlet of the SSR ascending path was designed to be small enough to prevent the rapid water level increase in the SSR. (authors)

Modifying Memory: Selectively Enhancing and Updating Personal Memories for a Museum Tour by Reactivating Them

Science.gov (United States)

St. Jacques, Peggy L.; Schacter, Daniel L.

2013-01-01

Memory can be modified when reactivated, but little is known about how the properties and extent of reactivation can selectively affect subsequent memory. We developed a novel museum paradigm to directly investigate reactivation-induced plasticity for personal memories. Participants reactivated memories triggered by photos taken from a camera they wore during a museum tour and made relatedness judgments on novel photos taken from a different tour of the same museum. Subsequent recognition memory for events at the museum was better for memories that were highly reactivated (i.e., the retrieval cues during reactivation matched the encoding experience) than for memories that were reactivated at a lower level (i.e., the retrieval cues during reactivation mismatched the encoding experience), but reactivation also increased false recognition of photographs depicting stops that were not experienced during the museum tour. Reactivation thus enables memories to be selectively enhanced and distorted via updating, thereby supporting the dynamic and flexible nature of memory. PMID:23406611

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.