Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

International Nuclear Information System (INIS)

Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

2001-03-01

JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

Analysis of incoloy 800ht alloy tested in thermal transient conditions

International Nuclear Information System (INIS)

Velciu, L.; Meleg, T.; Nitu, A.; Popa, L.

2015-01-01

This paper investigated Incoloy 800 HT alloy after following thermal transient tests: fast heating rates (50° and 90°C/minute) up to 1,000°C, maintaining this temperature level (0 and 60 minutes), furnace-cooling until 220°C, and then air-cooling. This alloy is one of the candidate materials for construction of the steam generators of the future NPP reactors. The analysis consisted in metallographic examination and traction tests. The samples were investigated using the Olympus GX 71 optical microscope, the OPL microdurometer with automatic cycle and WALTER BAI traction device. The average grain size was determined by linear interception method. The micro hardness was calculated by the relationship from the device technical book. On the traction diagrams were obtained: strength resistance (Rm), elongation at rupture (A) and elastic modulus (E). The tested alloy was compared with the ''as received'' material, and the results showed a good behavior of this alloy in the presented conditions. (authors)

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

Test system to simulate transient overpower LMFBR cladding failure

International Nuclear Information System (INIS)

Barrus, H.G.; Feigenbutz, L.V.

1981-01-01

One of the HEDL programs has the objective to experimentally characterize fuel pin cladding failure due to cladding rupture or ripping. A new test system has been developed which simulates a transient mechanically-loaded fuel pin failure. In this new system the mechanical load is prototypic of a fuel pellet rapidly expanding against the cladding due to various causes such as fuel thermal expansion, fuel melting, and fuel swelling. This new test system is called the Fuel Cladding Mechanical Interaction Mandrel Loading Test (FCMI/MLT). The FCMI/MLT test system and the method used to rupture cladding specimens very rapidly to simulate a transient event are described. Also described is the automatic data acquisition and control system which is required to control the startup, operation and shutdown of the very fast tests, and needed to acquire and store large quantities of data in a short time

40 CFR Appendix E to Subpart S of... - Transient Test Driving Cycle

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 2 2010-07-01 2010-07-01 false Transient Test Driving Cycle E Appendix... Driving Cycle (I) Driver's trace. All excursions in the transient driving cycle shall be evaluated by the... shall cause a test to be void. In addition, provisions shall be available to utilize cycle validation...

40 CFR 1048.510 - What transient duty cycles apply for laboratory testing?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false What transient duty cycles apply for... Procedures § 1048.510 What transient duty cycles apply for laboratory testing? (a) Starting with the 2007 model year, measure emissions by testing the engine on a dynamometer with the duty cycle described in...

Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

Energy Technology Data Exchange (ETDEWEB)

Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

2016-01-15

The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

International Nuclear Information System (INIS)

Roberto, Thiago D.; Silva, Mário A. B. da; Lapa, Celso M.F.

2016-01-01

The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

Electronically controlled mechanical seal for aerospace applications--Part 2: Transient tests

Science.gov (United States)

Wolff, Paul J.; Salant, Richard F.

1995-01-01

An electronically controlled mechanical seal for use as the purge gas seal in a liquid oxygen turbopump has been fabricated and tested under transient operating conditions. The thickness of the lubricating film is controlled by adjusting the coning of the carbon face. This is accomplished by applying a voltage to a piezoelectric actuator to which the carbon face is bonded. The seal has been operated with a closed-loop control system that utilizes either the leakage rate or seal face temperature as the feedback. Both speed and pressure transients have been imposed on the seal. The transient tests have demonstrated that the seal is capable of maintaing low leakage rates while limiting face temperatures.

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

International Nuclear Information System (INIS)

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

2004-07-01

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

40 CFR 1039.510 - Which duty cycles do I use for transient testing?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Which duty cycles do I use for... ENGINES Test Procedures § 1039.510 Which duty cycles do I use for transient testing? (a) Measure emissions by testing the engine on a dynamometer with one of the following transient duty cycles to determine...

Characterizing hydraulic fractures in shale gas reservoirs using transient pressure tests

Directory of Open Access Journals (Sweden)

Cong Wang

2015-06-01

This work presents an unconventional gas reservoir simulator and its application to quantify hydraulic fractures in shale gas reservoirs using transient pressure data. The numerical model incorporates most known physical processes for gas production from unconventional reservoirs, including two-phase flow of liquid and gas, Klinkenberg effect, non-Darcy flow, and nonlinear adsorption. In addition, the model is able to handle various types and scales of fractures or heterogeneity using continuum, discrete or hybrid modeling approaches under different well production conditions of varying rate or pressure. Our modeling studies indicate that the most sensitive parameter of hydraulic fractures to early transient gas flow through extremely low permeability rock is actually the fracture-matrix contacting area, generated by fracturing stimulation. Based on this observation, it is possible to use transient pressure testing data to estimate the area of fractures generated from fracturing operations. We will conduct a series of modeling studies and present a methodology using typical transient pressure responses, simulated by the numerical model, to estimate fracture areas created or to quantity hydraulic fractures with traditional well testing technology. The type curves of pressure transients from this study can be used to quantify hydraulic fractures in field application.

Transient testing of the FFTF for decay-heat removal by natural convection

International Nuclear Information System (INIS)

Beaver, T.R.; Johnson, H.G.; Stover, R.L.

1982-06-01

This paper reports on the series of transient tests performed in the FFTF as a major part of the pre-operations testing program. The structure of the transient test program was designed to verify the capability of the FFTF to safely remove decay heat by natural convection. The series culminated in a scram from full power to complete natural convection in the plant, simulating a loss of all electrical power. Test results and acceptance criteria related to the verification of safe decay heat removal are presented

Nuclear reactors transients identification and classification system; Sistema de identificacao e classificacao de transientes em reatores nucleares

Energy Technology Data Exchange (ETDEWEB)

Bianchi, Paulo Henrique

2008-07-01

This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)

Geometrical considerations in the transient ionization testing of digital logic circuits

International Nuclear Information System (INIS)

Johnston, A.

1982-01-01

Mechanisms are identified that can cause the transient response of digital logic circuits to depend on the logic state in which they are irradiated. Several of these mechanisms depend on surface topology, and for these cases the sensitive logic states can be determined by examining the topology. General approaches for transient radiation testing are also discussed for several MSI and LSI device technologies

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

Energy Technology Data Exchange (ETDEWEB)

Bradley K. Heath

2014-03-01

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

Nuclear reactors transients identification and classification system

International Nuclear Information System (INIS)

Bianchi, Paulo Henrique

2008-01-01

This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

International Nuclear Information System (INIS)

Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

2005-01-01

Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test

Behavior of mixed-oxide fuel elements during the TOPI-1E transient overpower test

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Yamamoto, K.; Hirai, K.; Shikakura, S.

1993-12-01

A slow-ramp, extended overpower transient test was conducted on a group of nineteen preirradiated mixed-oxide fuel elements in EBR-II. During the transient two of the test elements with high-density fuel and tempered martensitic cladding (PNC-FMS) breached at an overpower of ∼75%. Fuel elements with austenitic claddings (D9, PNC316, and PNC150), many with aggressive design features and high burnups, survived the overpower transient and incurred little or no cladding strain. Fuel elements with annual fuel or heterogeneous fuel columns also behaved well

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

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

International Nuclear Information System (INIS)

Meier, J.K.

1983-01-01

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

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

Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.

1994-10-01

From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s and the overpowers ranged between ∼60 and 100% of the elements' prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC's prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived

On the behaviour of dissolved fission gases prior to transient testing of fuel pins

International Nuclear Information System (INIS)

Wood, M.H.; Matthews, J.R.

1978-10-01

The TREAT and CABRI series of reactor safety experiments on irradiated fuel require the transfer of fuel pins from the reactor in which the fuel has achieved some burn-up to the test facility. Subsequently, the fuel is restored to power in the test facility for some time before transient heating is initiated. Such pre-test manoeuvres, where the fuel is subjected to changes in the fission rate and temperature, may have important consequences for the fission gas behaviour during the transient experiment. The results of rate theory calculations are used to assess these effects. (author)

PWR station blackout transient simulation in the INER integral system test facility

International Nuclear Information System (INIS)

Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

2004-01-01

Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Impacts Analyses Supporting the National Environmental Policy Act Environmental Assessment for the Resumption of Transient Testing Program

Energy Technology Data Exchange (ETDEWEB)

Annette L. Schafer; Lloyd C. Brown; David C. Carathers; Boyd D. Christensen; James J. Dahl; Mark L. Miller; Cathy Ottinger Farnum; Steven Peterson; A. Jeffrey Sondrup; Peter V. Subaiya; Daniel M. Wachs; Ruth F. Weiner

2013-11-01

Environmental and health impacts are presented for activities associated with transient testing of nuclear fuel and material using two candidate test reactors. Transient testing involves irradiation of nuclear fuel or materials for short time-periods under high neutron flux rates. The transient testing process includes transportation of nuclear fuel or materials inside a robust shipping cask to a hot cell, removal from the shipping cask, pre-irradiation examination of the nuclear materials, assembly of an experiment assembly, transportation of the experiment assembly to the test reactor, irradiation in the test reactor, transport back to the hot cell, and post-irradiation examination of the nuclear fuel or material. The potential for environmental or health consequences during the transportation, examination, and irradiation actions are assessed for normal operations, off-normal (accident) scenarios, and transportation. Impacts to the environment (air, soil, and groundwater), are assessed during each phase of the transient testing process. This report documents the evaluation of potential consequences to the general public. This document supports the Environmental Assessment (EA) required by the U.S. National Environmental Policy Act (NEPA) (42 USC Subsection 4321 et seq.).

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.

Behavior of metallic fuel in treat transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Wright, A.E.; Robinson, W.R.; Klickman, A.E.

1988-01-01

Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types

Increased Risk of Pregnancy Complications After Stroke : The FUTURE Study (Follow-Up of Transient Ischemic Attack and Stroke Patients and Unelucidated Risk Factor Evaluation)

NARCIS (Netherlands)

van Alebeek, Mayte E; de Vrijer, Myrthe; Arntz, Renate M; Maaijwee, Noortje A M M; Synhaeve, Nathalie E; Schoonderwaldt, Hennie; van der Vlugt, Maureen J; van Dijk, Ewoud J; de Heus, Roel; Rutten-Jacobs, Loes C A; de Leeuw, Frank-Erik

BACKGROUND AND PURPOSE: The study goal was to investigate the prevalence of pregnancy complications and pregnancy loss in women before, during, and after young ischemic stroke/transient ischemic attack. METHODS: In the FUTURE study (Follow-Up of Transient Ischemic Attack and Stroke Patients and

Modeling transient streaming potentials in falling-head permeameter tests.

Science.gov (United States)

Malama, Bwalya; Revil, André

2014-01-01

We present transient streaming potential data collected during falling-head permeameter tests performed on samples of two sands with different physical and chemical properties. The objective of the work is to estimate hydraulic conductivity (K) and the electrokinetic coupling coefficient (Cl ) of the sand samples. A semi-empirical model based on the falling-head permeameter flow model and electrokinetic coupling is used to analyze the streaming potential data and to estimate K and Cl . The values of K estimated from head data are used to validate the streaming potential method. Estimates of K from streaming potential data closely match those obtained from the associated head data, with less than 10% deviation. The electrokinetic coupling coefficient was estimated from streaming potential vs. (1) time and (2) head data for both sands. The results indicate that, within limits of experimental error, the values of Cl estimated by the two methods are essentially the same. The results of this work demonstrate that a temporal record of the streaming potential response in falling-head permeameter tests can be used to estimate both K and Cl . They further indicate the potential for using transient streaming potential data as a proxy for hydraulic head in hydrogeology applications. © 2013, National Ground Water Association.

Evaluation of the Non-Transient Hydrologic Source Term from the CAMBRIC Underground Nuclear Test in Frenchman Flat, Nevada Test Site

International Nuclear Information System (INIS)

Tompson, A B; Maxwell, R M; Carle, S F; Zavarin, M; Pawloski, G A.; Shumaker, D E

2005-01-01

Hydrologic Source Term (HST) calculations completed in 1998 at the CAMBRIC underground nuclear test site were LLNL's first attempt to simulate a hydrologic source term at the NTS by linking groundwater flow and transport modeling with geochemical modeling (Tompson et al., 1999). Significant effort was applied to develop a framework that modeled in detail the flow regime and captured all appropriate chemical processes that occurred over time. However, portions of the calculations were simplified because of data limitations and a perceived need for generalization of the results. For example: (1) Transient effects arising from a 16 years of pumping at the site for a radionuclide migration study were not incorporated. (2) Radionuclide fluxes across the water table, as derived from infiltration from a ditch to which pumping effluent was discharged, were not addressed. (3) Hydrothermal effects arising from residual heat of the test were not considered. (4) Background data on the ambient groundwater flow direction were uncertain and not represented. (5) Unclassified information on the Radiologic Source Term (RST) inventory, as tabulated recently by Bowen et al. (2001), was unavailable; instead, only a limited set of derived data were available (see Tompson et al., 1999). (6) Only a small number of radionuclides and geochemical reactions were incorporated in the work. (7) Data and interpretation of the RNM-2S multiple well aquifer test (MWAT) were not available. As a result, the current Transient CAMBRIC Hydrologic Source Term project was initiated as part of a broader Phase 2 Frenchman Flat CAU flow and transport modeling effort. The source term will be calculated under two scenarios: (1) A more specific representation of the transient flow and radionuclide release behavior at the site, reflecting the influence of the background hydraulic gradient, residual test heat, pumping experiment, and ditch recharge, and taking into account improved data sources and modeling

Barometric pressure transient testing applications at the Nevada Test Site: formation permeability analysis. Final report

International Nuclear Information System (INIS)

Hanson, J.M.

1984-12-01

The report evaluates previous investigations of the gas permeability of the rock surrounding emplacement holes at the Nevada Test Site. The discussion sets the framework from which the present uncertainty in gas permeability can be overcome. The usefulness of the barometric pressure testing method has been established. Flow models were used to evaluate barometric pressure transients taken at NTS holes U2fe, U19ac and U20ai. 31 refs., 103 figs., 18 tabs

Transient and steady-state tests of the space power research engine with resistive and motor loads

Science.gov (United States)

Rauch, Jeffrey S.; Kankam, M. David

1995-01-01

The NASA Lewis Research Center (LeRC) has been testing free-piston Stirling engine/linear alternators (FPSE/LA) to develop advanced power convertors for space-based electrical power generation. Tests reported herein were performed to evaluate the interaction and transient behavior of FPSE/LA-based power systems with typical user loads. Both resistive and small induction motor loads were tested with the space power research engine (SPRE) power system. Tests showed that the control system could maintain constant long term voltage and stable periodic operation over a large range of engine operating parameters and loads. Modest resistive load changes were shown to cause relatively large voltage and, therefore, piston and displacer amplitude excursions. Starting a typical small induction motor was shown to cause large and, in some cases, deleterious voltage transients. The tests identified the need for more effective controls, if FPSE/LAs are to be used for stand-alone power systems. The tests also generated a large body of transient dynamic data useful for analysis code validation.

Transient and Steady-state Tests of the Space Power Research Engine with Resistive and Motor Loads

Science.gov (United States)

Rauch, Jeffrey S.; Kankam, M. David

1995-01-01

The NASA Lewis Research Center (LeRC) has been testing free-piston Stirling engine/linear alternators (FPSE/LA) to develop advanced power convertors for space-based electrical power generation. Tests reported herein were performed to evaluate the interaction and transient behavior of FPSE/LA-based power systems with typical user loads. Both resistive and small induction motor loads were tested with the space power research engine (SPRE) power system. Tests showed that the control system could maintain constant long term voltage and stable periodic operation over a large range of engine operating parameters and loads. Modest resistive load changes were shown to cause relatively large voltage and, therefore, piston and displacer amplitude excursions. Starting a typical small induction motor was shown to cause large and, in some cases, deleterious voltage transients. The tests identified the need for more effective controls, if FPSE/LAs are to be used for stand-alone power systems. The tests also generated a large body of transient dynamic data useful for analysis code validation.

Testing the ontogenetic base for the transient model of inflorescence development.

Science.gov (United States)

Bull-Hereñu, Kester; Claßen-Bockhoff, Regine

2013-11-01

Current research in plant science has concentrated on revealing ontogenetic processes of key attributes in plant evolution. One recently discussed model is the 'transient model' successful in explaining some types of inflorescence architectures based on two main principles: the decline of the so called 'vegetativeness' (veg) factor and the transient nature of apical meristems in developing inflorescences. This study examines whether both principles find a concrete ontogenetic correlate in inflorescence development. To test the ontogenetic base of veg decline and the transient character of apical meristems the ontogeny of meristematic size in developing inflorescences was investigated under scanning electron microscopy. Early and late inflorescence meristems were measured and compared during inflorescence development in 13 eudicot species from 11 families. The initial size of the inflorescence meristem in closed inflorescences correlates with the number of nodes in the mature inflorescence. Conjunct compound inflorescences (panicles) show a constant decrease of meristematic size from early to late inflorescence meristems, while disjunct compound inflorescences present an enlargement by merging from early inflorescence meristems to late inflorescence meristems, implying a qualitative change of the apical meristems during ontogeny. Partial confirmation was found for the transient model for inflorescence architecture in the ontogeny: the initial size of the apical meristem in closed inflorescences is consistent with the postulated veg decline mechanism regulating the size of the inflorescence. However, the observed biphasic kinetics of the development of the apical meristem in compound racemes offers the primary explanation for their disjunct morphology, contrary to the putative exclusive transient mechanism in lateral axes as expected by the model.

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

Comparison of transient PCRV model test results with analysis

International Nuclear Information System (INIS)

Marchertas, A.H.; Belytschko, T.B.

1979-01-01

Comparisons are made of transient data derived from simple models of a reactor containment vessel with analytical solutions. This effort is a part of the ongoing process of development and testing of the DYNAPCON computer code. The test results used in these comparisons were obtained from scaled models of the British sodium cooled fast breeder program. The test structure is a scaled model of a cylindrically shaped reactor containment vessel made of concrete. This concrete vessel is prestressed axially by holddown bolts spanning the top and bottom slabs along the cylindrical walls, and is also prestressed circumferentially by a number of cables wrapped around the vessel. For test purposes this containment vessel is partially filled with water, which comes in direct contact with the vessel walls. The explosive charge is immersed in the pool of water and is centrally suspended from the top of the vessel. The load history was obtained from an ICECO analysis, using the equations of state for the source and the water. A detailed check of this solution was made to assure that the derived loading did provide the correct input. The DYNAPCON code was then used for the analysis of the prestressed concrete containment model. This analysis required the simulation of prestressing and the response of the model to the applied transient load. The calculations correctly predict the magnitudes of displacements of the PCRV model. In addition, the displacement time histories obtained from the calculations reproduce the general features of the experimental records: the period elongation and amplitude increase as compared to an elastic solution, and also the absence of permanent displacement. However, the period still underestimates the experiment, while the amplitude is generally somewhat large

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.

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.

Isolated transient vertigo: posterior circulation ischemia or benign origin?

Science.gov (United States)

Blasberg, Tobias F; Wolf, Lea; Henke, Christian; Lorenz, Matthias W

2017-06-14

Isolated transient vertigo can be the only symptom of posterior circulation ischemia. Thus, it is important to differentiate isolated vertigo of a cerebrovascular origin from that of more benign origins, as patients with cerebral ischemia have a much higher risk for future stroke than do those with 'peripheral' vertigo. The current study aims to identify risk factors for cerebrovascular origin of isolated transient vertigo, and for future cerebrovascular events. From the files of 339 outpatients with isolated transient vertigo we extracted history, clinical and technical findings, diagnosis, and follow-up information on subsequent stroke or transient ischemic attack (TIA). Risk factors were analyzed using multivariate regression models (logistic or Cox) and reconfirmed in univariate analyses. On first presentation, 48 (14.2%) patients received the diagnosis 'probable or definite cerebrovascular vertigo'. During follow-up, 41 patients suffered stroke or TIA (event rate 7.9 per 100 person years, 95% confidence interval (CI) 5.5-10.4), 26 in the posterior circulation (event rate 4.8 per 100 person years, 95% CI 3.0-6.7). The diagnosis was not associated with follow-up cerebrovascular events. In multivariate models testing multiple potential determinants, only the presentation mode was consistently associated with the diagnosis and stroke risk: patients who presented because of vertigo (rather than reporting vertigo when they presented for other reasons) had a significantly higher risk for future stroke or TIA (p = 0.028, event rate 13.4 vs. 5.4 per 100 person years) and for future posterior circulation stroke or TIA (p = 0.044, event rate 7.8 vs. 3.5 per 100 person years). We here report for the first time follow-up stroke rates in patients with transient isolated vertigo. In such patients, the identification of those with cerebrovascular origin remains difficult, and presentation mode was found to be the only consistent risk factor. Confirmation in an independent

Comparison of transient PCRV model test results with analysis

International Nuclear Information System (INIS)

Marchertas, A.H.; Belytschko, T.B.

1979-01-01

Comparisons are made of transient data derived from simple models of a reactor containment vessel with analytical solutions. This effort is a part of the ongoing process of development and testing of the DYNAPCON computer code. The test results used in these comparisons were obtained from scaled models of the British sodium cooled fast breeder program. The test structure is a scaled model of a cylindrically shaped reactor containment vessel made of concrete. This concrete vessel is prestressed axially by holddown bolts spanning the top and bottom slabs along the cylindrical walls, and is also prestressed circumferentially by a number of cables wrapped around the vessel. For test purposes this containment vessel is partially filled with water, which comes in direct contact with the vessel walls. The explosive charge is immersed in the pool of water and is centrally suspended from the top of the vessel. The tests are very similar to the series of tests made for the COVA experimental program, but the vessel here is the prestressed concrete container. (orig.)

ANO-2 turbine trip transient test analysis using MMS

International Nuclear Information System (INIS)

Jain, P.K.; Divakaruni, S.M.

1984-01-01

The data from the turbine trip transient tests conducted at the Arkansas Nuclear One-Unit 2 was used as one of the benchmark cases for validating the Modular Modeling System (MMS) Code, developed by the Electric Power Research Institute (EPRI). The data was used first to validate the modules in stand-alone simulation tests and then in a Nuclear Steam Supply system integral tests. This paper presents the results from the MMS simulation effort and compares the code generated results with the plant data as well as RETRAN results. In general, MMS simulation results compare very well with the plant data. The code calculations for the hot and cold leg temperatures, primary system pressure and the pressurizer level are very good compared to RETRAN; however, MMS results for steam generator level compare reasonably well only with RETRAN calculations

Impacts Analyses Supporting the National Environmental Policy Act Environmental Assessment for the Resumption of Transient Testing Program

Energy Technology Data Exchange (ETDEWEB)

Schafer, Annette L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Brown, LLoyd C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carathers, David C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Christensen, Boyd D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dahl, James J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Miller, Mark L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Farnum, Cathy Ottinger [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Peterson, Steven [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sondrup, A. Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Subaiya, Peter V. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wachs, Daniel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Weiner, Ruth F. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-02-01

This document contains the analysis details and summary of analyses conducted to evaluate the environmental impacts for the Resumption of Transient Fuel and Materials Testing Program. It provides an assessment of the impacts for the two action alternatives being evaluated in the environmental assessment. These alternatives are (1) resumption of transient testing using the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) and (2) conducting transient testing using the Annular Core Research Reactor (ACRR) at Sandia National Laboratory in New Mexico (SNL/NM). Analyses are provided for radiologic emissions, other air emissions, soil contamination, and groundwater contamination that could occur (1) during normal operations, (2) as a result of accidents in one of the facilities, and (3) during transport. It does not include an assessment of the biotic, cultural resources, waste generation, or other impacts that could result from the resumption of transient testing. Analyses were conducted by technical professionals at INL and SNL/NM as noted throughout this report. The analyses are based on bounding radionuclide inventories, with the same inventories used for test materials by both alternatives and different inventories for the TREAT Reactor and ACRR. An upper value on the number of tests was assumed, with a test frequency determined by the realistic turn-around times required between experiments. The estimates provided for impacts during normal operations are based on historical emission rates and projected usage rates; therefore, they are bounding. Estimated doses for members of the public, collocated workers, and facility workers that could be incurred as a result of an accident are very conservative. They do not credit safety systems or administrative procedures (such as evacuation plans or use of personal protective equipment) that could be used to limit worker doses. Doses estimated for transportation are conservative and are based on

Radiation-induced transient absorption in single mode optical fibers

International Nuclear Information System (INIS)

Looney, L.D.; Lyons, P.B.

1988-01-01

This paper reviews the measurements conducted by the Los Alamos National Laboratory in support of these NATO efforts wherein radiation-induced transient absorption was measured over time ranges from a few ns to several μs for two single mode fibers. Experimental conditions were varied to provide data for future development of standarized test conditions for single mode fibers. 8 refs., 11 figs

Wide Field Radio Transient Surveys

Science.gov (United States)

Bower, Geoffrey

2011-04-01

The time domain of the radio wavelength sky has been only sparsely explored. Nevertheless, serendipitous discovery and results from limited surveys indicate that there is much to be found on timescales from nanoseconds to years and at wavelengths from meters to millimeters. These observations have revealed unexpected phenomena such as rotating radio transients and coherent pulses from brown dwarfs. Additionally, archival studies have revealed an unknown class of radio transients without radio, optical, or high-energy hosts. The new generation of centimeter-wave radio telescopes such as the Allen Telescope Array (ATA) will exploit wide fields of view and flexible digital signal processing to systematically explore radio transient parameter space, as well as lay the scientific and technical foundation for the Square Kilometer Array. Known unknowns that will be the target of future transient surveys include orphan gamma-ray burst afterglows, radio supernovae, tidally-disrupted stars, flare stars, and magnetars. While probing the variable sky, these surveys will also provide unprecedented information on the static radio sky. I will present results from three large ATA surveys (the Fly's Eye survey, the ATA Twenty CM Survey (ATATS), and the Pi GHz Survey (PiGSS)) and several small ATA transient searches. Finally, I will discuss the landscape and opportunities for future instruments at centimeter wavelengths.

Transient performance of EBR-II driver fuel

International Nuclear Information System (INIS)

Buzzell, J.A.; Hudman, G.D.; Porter, D.L.

1981-01-01

The first phases of qualification of the EBR-II driver fuel for repeated transient overpower operation have recently been completed. The accomplishments include prediction of the transient fuel and cladding performance through ex-core testing and fuel-element modeling studies, localized in-core power testing during steady-state operation, and whole-core multiple transient testing. The metallic driver fuel successfully survived 56 transients, spaced over a 45-day period, with power increases of approx. 160% at rates of approx. 1%/s with a 720-second hold at full power. The performance results obtained from both ex-core and n-core tests indicate that the fuel is capable of repeated transient operation

Test programmes of HTTR for future generation HTGRs

International Nuclear Information System (INIS)

Kunitomi, K.; Tachibana, Y.; Takeda, T.; Saikusa, A.; Shiozawa, S.

1997-01-01

Test programs of the High Temperature Engineering Test Reactor (HTTR) for future generation High Temperature Gas-Cooled Reactors (HTGRs) have been established considering design and development status of HTGRs in Japan and the world. Test programs are divided into six categories, thermal hydraulics, fuel, safety, high temperature components, core physics and control-instrumentations. All programs are related to the technology of future generation HTGRs and will be submitted to a new Coordinated Research Program (CRP) so that all participants from the world in test programs of the HTTR can use measured data for their future generation.HTGRs. This paper describes test programs of the HTTR for the development of future generation HTGRs after explanation of a future generation HTGR in Japan. (author)

A follow-up of transients. Stage 1

International Nuclear Information System (INIS)

Enekull, Aa.; Wallner, B.

1981-09-01

A follow-up of the transients of temperature and pressure in the primary pressurized system of a nuclear power plant has been completed for the Barsebaeck-1 reactor. The investigation consists of the following steps:- the collation of transients - drawing up load data based on transients-analyses of stress - recommendations for future programs. It was found that the lifetime of the system will exceed 40 years excluding some of the pipes for feed water. The appendices give a detailed description of the transients.(G.B.)

The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

1990-01-01

This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

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

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

Have the Software Testing a Future?

Directory of Open Access Journals (Sweden)

Juan A. Godoy

2012-06-01

Full Text Available Software testing is directed to a dark future, with greater political isolation management, less funding and poorer overall quality. The hopes of the theory of software quality and test new technologies of the 1990s have been usurped by "tastes" in the development focused on ideas such as "Agile", "Object Oriented", "Cloud” and applications “Mobile” of $ 0.99. The new languages and development methods are designed to allow developers to "throw" code faster and not to improve versions, maintenance, testing and traceability or auditing. The costs of maintenance and development will increase, the budgets for the test will fall and more projects fail. The future of the tests is shade. In this article is analyzed this situation.

Vivitron 1995, transient voltage simulation, high voltage insulator tests, electric field calculation

International Nuclear Information System (INIS)

Frick, G.; Osswald, F.; Heusch, B.

1996-01-01

Preliminary investigations showed clearly that, because of the discrete electrode structure of the Vivitron, important overvoltage leading to insulator damage can appear in case of a spark. The first high voltage tests showed damage connected with such events. This fact leads to a severe voltage limitation. This work describes, at first, studies made to understand the effects of transients and the associated over-voltage appearing in the Vivitron. Then we present the high voltage tests made with full size Vivitron components using the CN 6 MV machine as a pilot machine. Extensive field calculations were made. These involve simulations of static stresses and transient overvoltages, on insulating boards and electrodes. This work gave us the solutions for arrangements and modifications in the machine. After application, the Vivitron runs now without any sparks and damage at 20 MV. In the same manner, we tested column insulators of a new design and so we will find out how to get to higher voltages. Electric field calculation around the tie bars connecting the discrete electrodes together showed field enhancements when the voltages applied on the discrete electrodes are not equally distributed. This fact is one of the sources of discharges and voltage limitations. A scenario of a spark event is described and indications are given how to proceed towards higher voltages, in the 30 MV range. (orig.)

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

Sextant: an expert system for transient analysis of nuclear reactors and integral test facilities

International Nuclear Information System (INIS)

Barbet, N.; Dumas, M.; Mihelich, G.

1987-01-01

Expert systems provide a new way of dealing with the computer-aided management of nuclear plants by combining several knowledge bases and reasoning modes together with a set of numerical models for real-time analysis of transients. New development tools are required together with metaknowledge bases handling temporal hypothetical reasoning and planning. They have to be efficient and robust because during a transient, neither measurements nor models, nor scenarios are hold as absolute references. SEXTANT is a general purpose physical analyzer intended to provide a pattern and avoid duplication of general tools and knowledge bases for similar applications. It combines several knowledge bases concerning measurements, models and qualitative behavior of PWR with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. For its development, SEXTANT requires a powerful shell. SPIRAL is such a toolkit, oriented towards online analysis of complex processes and already used in several applications

Classification methods for noise transients in advanced gravitational-wave detectors II: performance tests on Advanced LIGO data

International Nuclear Information System (INIS)

Powell, Jade; Heng, Ik Siong; Torres-Forné, Alejandro; Font, José A; Lynch, Ryan; Trifirò, Daniele; Cuoco, Elena; Cavaglià, Marco

2017-01-01

The data taken by the advanced LIGO and Virgo gravitational-wave detectors contains short duration noise transients that limit the significance of astrophysical detections and reduce the duty cycle of the instruments. As the advanced detectors are reaching sensitivity levels that allow for multiple detections of astrophysical gravitational-wave sources it is crucial to achieve a fast and accurate characterization of non-astrophysical transient noise shortly after it occurs in the detectors. Previously we presented three methods for the classification of transient noise sources. They are Principal Component Analysis for Transients (PCAT), Principal Component LALInference Burst (PC-LIB) and Wavelet Detection Filter with Machine Learning (WDF-ML). In this study we carry out the first performance tests of these algorithms on gravitational-wave data from the Advanced LIGO detectors. We use the data taken between the 3rd of June 2015 and the 14th of June 2015 during the 7th engineering run (ER7), and outline the improvements made to increase the performance and lower the latency of the algorithms on real data. This work provides an important test for understanding the performance of these methods on real, non stationary data in preparation for the second advanced gravitational-wave detector observation run, planned for later this year. We show that all methods can classify transients in non stationary data with a high level of accuracy and show the benefits of using multiple classifiers. (paper)

Coincident steam generator tube rupture and stuck-open safety relief valve carryover tests: MB-2 steam generator transient response test program

International Nuclear Information System (INIS)

Garbett, K.; Mendler, O.J.; Gardner, G.C.; Garnsey, R.; Young, M.Y.

1987-03-01

In PWR steam generator tube rupture (SGTR) faults, a direct pathway for the release of radioactive fission products can exist if there is a coincident stuck-open safety relief valve (SORV) or if the safety relief valve is cycled. In addition to the release of fission products from the bulk steam generator water by moisture carryover, there exists the possibility that some primary coolant may be released without having first mixed with the bulk water - a process called primary coolant bypassing. The MB-2 Phase II test program was designed specifically to identify the processes for droplet carryover during SGTR faults and to provide data of sufficient accuracy for use in developing physical models and computer codes to describe activity release. The test program consisted of sixteen separate tests designed to cover a range of steady-state and transient fault conditions. These included a full SGTR/SORV transient simulation, two SGTR overfill tests, ten steady-state SGTR tests at water levels ranging from very low levels in the bundle up to those when the dryer was flooded, and three moisture carryover tests without SGTR. In these tests the influence of break location and the effect of bypassing the dryer were also studied. In a final test the behavior with respect to aerosol particles in a dry steam generator, appropriate to a severe accident fault, was investigated

Transient productivity index for numerical well test simulations

Energy Technology Data Exchange (ETDEWEB)

Blanc, G.; Ding, D.Y.; Ene, A. [Institut Francais du Petrole, Pau (France)] [and others

1997-08-01

The most difficult aspect of numerical simulation of well tests is the treatment of the Bottom Hole Flowing (BHF) Pressure. In full field simulations, this pressure is derived from the Well-block Pressure (WBP) using a numerical productivity index which accounts for the grid size and permeability, and for the well completion. This productivity index is calculated assuming a pseudo-steady state flow regime in the vicinity of the well and is therefore constant during the well production period. Such a pseudo-steady state assumption is no longer valid for the early time of a well test simulation as long as the pressure perturbation has not reached several grid-blocks around the well. This paper offers two different solutions to this problem: (1) The first one is based on the derivation of a Numerical Transient Productivity Index (NTPI) to be applied to Cartesian grids; (2) The second one is based on the use of a Corrected Transmissibility and Accumulation Term (CTAT) in the flow equation. The representation of the pressure behavior given by both solutions is far more accurate than the conventional one as shown by several validation examples which are presented in the following pages.

Fuel cladding mechanical properties for transient analysis

International Nuclear Information System (INIS)

Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

1976-01-01

Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

Energy Technology Data Exchange (ETDEWEB)

Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); deHart, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-07-11

This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/k$. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$_2$, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the

Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

International Nuclear Information System (INIS)

Ortensi, Javier; Baker, Benjamin; Wang, Yaqi; Schunert, Sebastian; DeHart, Mark

2017-01-01

This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/k$. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$ 2 $, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the

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

Comparison of in-plant performance test data with analytic prediction of reactor safety system injection transient (U)

International Nuclear Information System (INIS)

Roy, B.N.; Neill, C.H. Jr.

1993-01-01

This paper compares the performance test data from injection transients for both of the subsystems of the Supplementary Safety System of the Savannah River Site production reactor with analytical predictions from an in-house thermal hydraulic computer code. The code was initially developed for design validation of the new Supplementary Safety System subsystem, but is shown to be equally capable of predicting the performance of the Supplementary Safety System existing subsystem even though the two subsystem transient injections have marked differences. The code itself was discussed and its validation using prototypic tests with simulated fluids was reported in an earlier paper (Roy and Nomm 1991)

Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

1986-01-01

TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

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.

First TREAT transient overpower tests on U-Pu-Zr fuel: M5 and M6

International Nuclear Information System (INIS)

Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

1987-01-01

Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) test of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the integral fast reactor concept. Similar tests on U-5 wt% Fs fueled Experimental Breeder Reactor (EBR)-II driver pins were previously performed and reported. Results from these earlier tests indicated a margin to failure of ∼ 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuels types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety-related fuel performance characteristics of U-Fs and U-Pu-Zr. Tests M5 and M6 indicate that, under the TOP conditions used in the tests, ternary fuel displayed about the same margin to failure as U-Fs fuel. At low burnups, ternary fuel showed less prefailure axial elongation than observed in U-Fs pins, but elongations of 3 to 5% might turn out to be typical. Finally, fuel from the breached ternary pin in M6 showed, qualitatively, the same benignly dispersive behavior as U-Fs

Axial location of cladding failure during a slow transient overpower TREAT test

International Nuclear Information System (INIS)

Page, R.J.; Murphy, W.F.; Holland, J.W.

1983-01-01

The axial location of cladding failure following a transient overpower accident is of importance in fast reactor safety studies in that it is a determining factor in the relocation of fuel, and therefore in the possibility of inherent neutronic shutdown of the reactor. In-pile experimental data on the axial location of cladding failure of fuel in bundles of pins is sparse since, in general, the experimental fuel pin bundles are largely destroyed during the in-pile test. The post-test examination work has been completed for TREAT test J1. It was found that damage to the fuel elements during the irradiation was low enough for an accurate observation of the location of cladding failure to be made for each of the seven pins

Numerical Analysis of S-CO{sub 2} Test Loop Transient Conditions near the Critical Point of CO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Bae, Seong Jun; Oh, Bongseong; Ahn, Yoonhan; Baik, Seongjoon; Lee, Jekyoung; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

It was identified that controlling CO{sub 2} compressor operation near the critical point is one of the most important issues to operate a S-CO{sub 2} Brayton cycle with a high efficiency. Despite the growing interest in the S-CO{sub 2} Brayton cycle, a few previous research on the transient analysis of the S-CO{sub 2} system has been conducted previously. Moreover, previous studies have some limitation in the modelled test facility, and the experiment was not performed to observe specific scenario. The KAIST research team has conducted S-CO{sub 2} system transient experiments with the CO{sub 2} compressing test facility called SCO{sub 2}PE (Supercritical CO{sub 2} Pressurizing Experiment) at KAIST In this study, authors use the transient analysis code GAMMA (Gas Multidimensional Multicomponent mixture Analysis) code for analyzing the experiment. Two transient scenarios were selected in this study; over cooling and under cooling situations. The selected transient situation is of particular interest since the compressor inlet conditions start to drift away from the critical point of CO{sub 2}. The results represent that the GAMMA code can simulate the S-CO{sub 2} test facility, SCO{sub 2}PE. However, as shown in the cooling water flow rate increasing scenario, the GAMMA code shows calculation error when the phase change occurs. Furthermore, although the results of the cooling water flow rate decrease case shows reasonable agreement with the experimental data, there are still some unexplained differences between the experimental data and the GAMMA code prediction.

Transient analysis for PWR reactor core using neural networks predictors

International Nuclear Information System (INIS)

Gueray, B.S.

2001-01-01

In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

RETRAN sensitivity studies of light water reactor transients. Final report

International Nuclear Information System (INIS)

Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.

1977-06-01

This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development

UNSUPERVISED TRANSIENT LIGHT CURVE ANALYSIS VIA HIERARCHICAL BAYESIAN INFERENCE

International Nuclear Information System (INIS)

Sanders, N. E.; Soderberg, A. M.; Betancourt, M.

2015-01-01

Historically, light curve studies of supernovae (SNe) and other transient classes have focused on individual objects with copious and high signal-to-noise observations. In the nascent era of wide field transient searches, objects with detailed observations are decreasing as a fraction of the overall known SN population, and this strategy sacrifices the majority of the information contained in the data about the underlying population of transients. A population level modeling approach, simultaneously fitting all available observations of objects in a transient sub-class of interest, fully mines the data to infer the properties of the population and avoids certain systematic biases. We present a novel hierarchical Bayesian statistical model for population level modeling of transient light curves, and discuss its implementation using an efficient Hamiltonian Monte Carlo technique. As a test case, we apply this model to the Type IIP SN sample from the Pan-STARRS1 Medium Deep Survey, consisting of 18,837 photometric observations of 76 SNe, corresponding to a joint posterior distribution with 9176 parameters under our model. Our hierarchical model fits provide improved constraints on light curve parameters relevant to the physical properties of their progenitor stars relative to modeling individual light curves alone. Moreover, we directly evaluate the probability for occurrence rates of unseen light curve characteristics from the model hyperparameters, addressing observational biases in survey methodology. We view this modeling framework as an unsupervised machine learning technique with the ability to maximize scientific returns from data to be collected by future wide field transient searches like LSST

UNSUPERVISED TRANSIENT LIGHT CURVE ANALYSIS VIA HIERARCHICAL BAYESIAN INFERENCE

Energy Technology Data Exchange (ETDEWEB)

Sanders, N. E.; Soderberg, A. M. [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Betancourt, M., E-mail: nsanders@cfa.harvard.edu [Department of Statistics, University of Warwick, Coventry CV4 7AL (United Kingdom)

2015-02-10

Historically, light curve studies of supernovae (SNe) and other transient classes have focused on individual objects with copious and high signal-to-noise observations. In the nascent era of wide field transient searches, objects with detailed observations are decreasing as a fraction of the overall known SN population, and this strategy sacrifices the majority of the information contained in the data about the underlying population of transients. A population level modeling approach, simultaneously fitting all available observations of objects in a transient sub-class of interest, fully mines the data to infer the properties of the population and avoids certain systematic biases. We present a novel hierarchical Bayesian statistical model for population level modeling of transient light curves, and discuss its implementation using an efficient Hamiltonian Monte Carlo technique. As a test case, we apply this model to the Type IIP SN sample from the Pan-STARRS1 Medium Deep Survey, consisting of 18,837 photometric observations of 76 SNe, corresponding to a joint posterior distribution with 9176 parameters under our model. Our hierarchical model fits provide improved constraints on light curve parameters relevant to the physical properties of their progenitor stars relative to modeling individual light curves alone. Moreover, we directly evaluate the probability for occurrence rates of unseen light curve characteristics from the model hyperparameters, addressing observational biases in survey methodology. We view this modeling framework as an unsupervised machine learning technique with the ability to maximize scientific returns from data to be collected by future wide field transient searches like LSST.

Tightly coupled transient analysis of EBR-II

International Nuclear Information System (INIS)

Makowitz, H.; Lehto, W.K.; Sackett, J.I.

1988-01-01

A Tightly Coupled transient analysis system for the Experimental Breeder Reactor-II (EBR-II) is currently being tested. The system consists of a faster than real time high fidelity reactor simulation, advanced graphics displays, expert system coupling, and real time data coupling via the EBR-II data acquisition system to and from the plant and the control system. The base, first generation software has been developed and is presently being tested. Various subsystem couplings and the total system integration are being checked out. This system should enhance the diagnostic and prognostic capability of EBR-II in the near term and provide automatic control during startup and power maneuvering in the future, as well as serve as a testbed for new control system development for advanced reactors

A transient overpower experiment in EBR-II

International Nuclear Information System (INIS)

Herzog, J.P.; Tsai, H.; Dean, E.M.; Aoyama, T.; Yamamoto, K.

1994-01-01

The TOPI-IE test was a transient overpower test on irradiate mixed-oxide fuel pins in the Experimental Breeder Reactor-II (EBR-II). The test, the fifth in a series, was part of a cooperative program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan to conduct operational transient testing on mixed-oxide fuel pins in the metal-fueled EBR-II. The principle objective of the TOPI-1E test was to assess breaching margins for irradiated mixed-oxide fuel pins over the Plant Protection System (PPS) thresholds during a slow, extended overpower transient. This paper describes the effect of the TOPI-1E experiment on reactor components and the impact of the experiment on the long-term operability of the reactor. The paper discusses the role that SASSYS played in the pre-test safety analysis of the experiment. The ability of SASSYS to model transient overpower events is detailed by comparisons of data from the experiment with computed reactor variables from a SASSYS post-test simulation of the experiment

Simulation of the transient behaviour of the Mulheim-Karlich nuclear power station and comparison with startup tests

International Nuclear Information System (INIS)

Burchhardt, W.; Weber, J.P.

1987-01-01

Among the startup tests of the Mulheim-Karlich nuclear power station an anticipated turbine trip and a rector trip have been performed. Calculations have been done for both tests by the authors using the computer codes DYNODE-P and ALMOD4. At a given reactor power the transient behavior of the plant is determined by the heat load of the once-through steam generators (OTSG), which in turn is governed by the integral control system. As expected the behavior of the plant after the trips can be exactly reproduced only if the bypass control system is completely modelled. In addition the heat transfer to the secondary side of the OTSG which governs the steam production has to be calculated close to reality. comparing the calculated results with the measured test data it is demonstrated that either code can be used to reproduce the transient behavior of the plant

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

International Nuclear Information System (INIS)

Wambach, J.L.

1980-01-01

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

A study of the transient performance of hydrostatic journal bearings. I - Test apparatus and facility

Science.gov (United States)

Scharrer, J. K.; Tellier, J.; Hibbs, R.

1992-10-01

A test apparatus was developed for studies of the transient performance of hydrostatic journal bearings operating in liquid nitrogen. The data obtained give the number of revolutions of the shaft contact before the liftoff and after touchdown as a function of bearing/shaft material combinations and operating conditions.

The OECD/NEA/NSC PBMR 400 MW coupled neutronics thermal hydraulics transient benchmark: transient results - 290

International Nuclear Information System (INIS)

Strydom, G.; Reitsma, F.; Ngeleka, P.T.; Ivanov, K.N.

2010-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well defined multi-dimensional computational benchmark problems with a common given set of cross sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified. (authors)

Recent development of transient electronics

Directory of Open Access Journals (Sweden)

Huanyu Cheng

2016-01-01

Full Text Available Transient electronics are an emerging class of electronics with the unique characteristic to completely dissolve within a programmed period of time. Since no harmful byproducts are released, these electronics can be used in the human body as a diagnostic tool, for instance, or they can be used as environmentally friendly alternatives to existing electronics which disintegrate when exposed to water. Thus, the most crucial aspect of transient electronics is their ability to disintegrate in a practical manner and a review of the literature on this topic is essential for understanding the current capabilities of transient electronics and areas of future research. In the past, only partial dissolution of transient electronics was possible, however, total dissolution has been achieved with a recent discovery that silicon nanomembrane undergoes hydrolysis. The use of single- and multi-layered structures has also been explored as a way to extend the lifetime of the electronics. Analytical models have been developed to study the dissolution of various functional materials as well as the devices constructed from this set of functional materials and these models prove to be useful in the design of the transient electronics.

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

Heat and mass release for some transient fuel source fires: A test report

International Nuclear Information System (INIS)

Nowlen, S.P.

1986-10-01

Nine fire tests using five different trash fuel source packages were conducted by Sandia National Laboratories. This report presents the findings of these tests. Data reported includes heat and mass release rates, total heat and mass release, plume temperatures, and average fuel heat of combustion. These tests were conducted as a part of the US Nuclear Regulatory Commission sponsored fire safety research program. Data from these tests were intended for use in nuclear power plant probabilistic risk assessment fire analyses. The results were also used as input to a fire test program at Sandia investigating the vulnerability of electrical control cabinets to fire. The fuel packages tested were chosen to be representative of small to moderately sized transient trash fuel sources of the type that would be found in a nuclear power plant. The highest fire intensity encountered during these tests was 145 kW. Plume temperatures did not exceed 820 0 C

Advanced test reactor testing experience-past, present and future

International Nuclear Information System (INIS)

Marshall, Frances M.

2006-01-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

EBR-II high-ramp transients under computer control

International Nuclear Information System (INIS)

Forrester, R.J.; Larson, H.A.; Christensen, L.J.; Booty, W.F.; Dean, E.M.

1983-01-01

During reactor run 122, EBR-II was subjected to 13 computer-controlled overpower transients at ramps of 4 MWt/s to qualify the facility and fuel for transient testing of LMFBR oxide fuels as part of the EBR-II operational-reliability-testing (ORT) program. A computer-controlled automatic control-rod drive system (ACRDS), designed by EBR-II personnel, permitted automatic control on demand power during the transients

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.

Characterizing SI Engine Transient Fuel Consumption in ALPHA

Science.gov (United States)

Examine typical transient engine operation encountered over the EPA's vehicle and engine testing drive cycles to characterize that transient fuel usage, and then describe the changes made to ALPHA to better model transient engine operation.

Models for transient analyses in advanced test reactors

International Nuclear Information System (INIS)

Gabrielli, Fabrizio

2011-01-01

Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are applied

Search for Transient Gravitational Waves in Coincidence with Short-Duration Radio Transients During 2007-2013

Science.gov (United States)

Abbott, B. P.; Hughey, Brennan; Zanolin, Michele; Szczepanczyk, Marek; Gill, Kiranjyot; Abbott, R.; Abbott, T. D.; Abernathy, M. R.; Acernese, F.; Ackley, K.;

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

Comparison of a fission-gas effects in a transient overpower test (HUT 5-7A) to FRAS3 code predictions

International Nuclear Information System (INIS)

Gruber, E.E.; Randklev, E.H.

1979-01-01

Fission gas has an important bearing on fuel dynamics during reactor transients. Fission-gas bubble sizes and densities, both within grains and on grain boundaries, are characterized as functions of radial location at the axial midplane in studies of PNL-9 fuel microstructures before and after the HUT 5-7A (PNL 9-25) TREAT test. The FRAS3 code, being developed to model fission-gas effects in reactor transients, is applied to analyze the results of the HUT 5-7A test are presented to illustrate the observed phenomena and the validity of the modeling approach

Tucurui`s 500 kV S F{sub 6} gas-insulated substation modeling for fast transients - field tests comparison

Energy Technology Data Exchange (ETDEWEB)

Lima Filho, Joao Ferreira de; Pazo Blanco, Flavio A.F.; D` Ajuz, Ary; Nascimento, Lidio F.A. [ELETRONORTE, Brasilia, DF (Brazil)

1994-12-31

Many problems in insulation, devices and equipment have been resulted from switching operations in the 500 kV Gas-Insulated Substation (GIS) of Tucurui power plant. Disconnector closing or opening action can subject the GIS components to great stresses from excessive voltages at high frequencies caused by successive reflections of travelling waves at discontinuities (1,2,3). Field tests were carried out in September 18-23, 1991 to verify the actual GIS stresses at Tucurui. From these results simulations on the Alternative Transients Program (ATP) were performed in order to check the substation modeling. This paper presents and discusses the Tucurui GIS modeling for switching fast transients and its field test validation. (author) 8 refs., 6 figs.

Volume 1. 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

A comprehensive compilation of the current (January 1975) state-of-the-art understanding of fuel pin response to reactivity insertion events up to and including failure threshold is presented. All mixed oxide fuel, transient overpower tests conducted to date are described and the data collected therein are presented. Current understanding of fuel pin cladding failure mechanisms is discussed, and interpretations of fuel pin cladding failure thresholds are presented as revealed by individual tests and by the several tests collectively. Plans for future tests are discussed

Severe transient tests on operation steam generators: Analysis of the fluid structure dynamic thermal interaction

International Nuclear Information System (INIS)

Billon, F.; David, J.; Procaccia, H.

1983-01-01

The operating efficiency of steam generators (S.G.s) and their structural integrity depend on the design configurations of the feedwater spray within the S.G., and on the operating procedure. To check the merit of some design modifications, and to verify the fluid-structure interaction with a view to preserve the S.G.s integrity during severe operating transients, a special instrumentation that admits the determination of the instantaneous thermal hydraulic characteristics of the flow in the secondary water and the S.G. tube sheet, has been installed by EDF on one steam generator of Tricastin unit 1 power plant. In parallel, FRAMATOME has developped a computer code, TEMPTRON, that allows the calculations of the thermal loads and the consequent stresses in the most sollicited zones of the steam generator during transient operation of the plant. This code divides the S.G. into three parts: - the first concerns the S.G.s region above the downcomer, zone where the mixing between hot water and cold feedwater occurs, - the second is the downcomer itself which is divided into n segments, - the third concerns the tube sheet zone which is also divided into n segments. The most severe transient test performed is the auxiliary cold feedwater injection into the steam generator during a hot standby of the plant: two levels of flow rate have been realised: 55 and 110 m 3 /h of 42 0 C feedwater. The tests have shown that if the cold feedwater injection occurs when the steam generator water level is below feedwater ring, the lowest fluid temperature reached at tube sheet inlet is about 230 0 C. (orig.)

Application of transient analysis methodology to heat exchanger performance monitoring

International Nuclear Information System (INIS)

Rampall, I.; Soler, A.I.; Singh, K.P.; Scott, B.H.

1994-01-01

A transient testing technique is developed to evaluate the thermal performance of industrial scale heat exchangers. A Galerkin-based numerical method with a choice of spectral basis elements to account for spatial temperature variations in heat exchangers is developed to solve the transient heat exchanger model equations. Testing a heat exchanger in the transient state may be the only viable alternative where conventional steady state testing procedures are impossible or infeasible. For example, this methodology is particularly suited to the determination of fouling levels in component cooling water system heat exchangers in nuclear power plants. The heat load on these so-called component coolers under steady state conditions is too small to permit meaningful testing. An adequate heat load develops immediately after a reactor shutdown when the exchanger inlet temperatures are highly time-dependent. The application of the analysis methodology is illustrated herein with reference to an in-situ transient testing carried out at a nuclear power plant. The method, however, is applicable to any transient testing application

Statistical searches for microlensing events in large, non-uniformly sampled time-domain surveys: A test using palomar transient factory data

Energy Technology Data Exchange (ETDEWEB)

Price-Whelan, Adrian M.; Agüeros, Marcel A. [Department of Astronomy, Columbia University, 550 W 120th Street, New York, NY 10027 (United States); Fournier, Amanda P. [Department of Physics, Broida Hall, University of California, Santa Barbara, CA 93106 (United States); Street, Rachel [Las Cumbres Observatory Global Telescope Network, Inc., 6740 Cortona Drive, Suite 102, Santa Barbara, CA 93117 (United States); Ofek, Eran O. [Benoziyo Center for Astrophysics, Weizmann Institute of Science, 76100 Rehovot (Israel); Covey, Kevin R. [Lowell Observatory, 1400 West Mars Hill Road, Flagstaff, AZ 86001 (United States); Levitan, David; Sesar, Branimir [Division of Physics, Mathematics, and Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States); Laher, Russ R.; Surace, Jason, E-mail: adrn@astro.columbia.edu [Spitzer Science Center, California Institute of Technology, Mail Stop 314-6, Pasadena, CA 91125 (United States)

2014-01-20

Many photometric time-domain surveys are driven by specific goals, such as searches for supernovae or transiting exoplanets, which set the cadence with which fields are re-imaged. In the case of the Palomar Transient Factory (PTF), several sub-surveys are conducted in parallel, leading to non-uniform sampling over its ∼20,000 deg{sup 2} footprint. While the median 7.26 deg{sup 2} PTF field has been imaged ∼40 times in the R band, ∼2300 deg{sup 2} have been observed >100 times. We use PTF data to study the trade off between searching for microlensing events in a survey whose footprint is much larger than that of typical microlensing searches, but with far-from-optimal time sampling. To examine the probability that microlensing events can be recovered in these data, we test statistics used on uniformly sampled data to identify variables and transients. We find that the von Neumann ratio performs best for identifying simulated microlensing events in our data. We develop a selection method using this statistic and apply it to data from fields with >10 R-band observations, 1.1 × 10{sup 9} light curves, uncovering three candidate microlensing events. We lack simultaneous, multi-color photometry to confirm these as microlensing events. However, their number is consistent with predictions for the event rate in the PTF footprint over the survey's three years of operations, as estimated from near-field microlensing models. This work can help constrain all-sky event rate predictions and tests microlensing signal recovery in large data sets, which will be useful to future time-domain surveys, such as that planned with the Large Synoptic Survey Telescope.

Multibeam Gpu Transient Pipeline for the Medicina BEST-2 Array

Science.gov (United States)

Magro, A.; Hickish, J.; Adami, K. Z.

2013-09-01

Radio transient discovery using next generation radio telescopes will pose several digital signal processing and data transfer challenges, requiring specialized high-performance backends. Several accelerator technologies are being considered as prototyping platforms, including Graphics Processing Units (GPUs). In this paper we present a real-time pipeline prototype capable of processing multiple beams concurrently, performing Radio Frequency Interference (RFI) rejection through thresholding, correcting for the delay in signal arrival times across the frequency band using brute-force dedispersion, event detection and clustering, and finally candidate filtering, with the capability of persisting data buffers containing interesting signals to disk. This setup was deployed at the BEST-2 SKA pathfinder in Medicina, Italy, where several benchmarks and test observations of astrophysical transients were conducted. These tests show that on the deployed hardware eight 20 MHz beams can be processed simultaneously for 640 Dispersion Measure (DM) values. Furthermore, the clustering and candidate filtering algorithms employed prove to be good candidates for online event detection techniques. The number of beams which can be processed increases proportionally to the number of servers deployed and number of GPUs, making it a viable architecture for current and future radio telescopes.

Transient dynamic finite element analysis of hydrogen distribution test chamber structure for hydrogen combustion loads

International Nuclear Information System (INIS)

Singh, R.K.; Redlinger, R.; Breitung, W.

2005-09-01

Design and analysis of blast resistant structures is an important area of safety research in nuclear, aerospace, chemical process and vehicle industries. Institute for Nuclear and Energy Technologies (IKET) of Research Centre- Karlsruhe (Forschungszentrum Karlsruhe or FZK) in Germany is pursuing active research on the entire spectrum of safety evaluation for efficient hydrogen management in case of the postulated design basis and beyond the design basis severe accidents for nuclear and non-nuclear applications. This report concentrates on the consequence analysis of hydrogen combustion accidents with emphasis on the structural safety assessment. The transient finite element simulation results obtained for 2gm, 4gm, 8gm and 16gm hydrogen combustion experiments concluded recently on the test-cell structure are described. The frequencies and damping of the test-cell observed during the hammer tests and the combustion experiments are used for the present three dimensional finite element model qualification. For the numerical transient dynamic evaluation of the test-cell structure, the pressure time history data computed with CFD code COM-3D is used for the four combustion experiments. Detail comparisons of the present numerical results for the four combustion experiments with the observed time signals are carried out to evaluate the structural connection behavior. For all the combustion experiments excellent agreement is noted for the computed accelerations and displacements at the standard transducer locations, where the measurements were made during the different combustion tests. In addition inelastic analysis is also presented for the test-cell structure to evaluate the limiting impulsive and quasi-static pressure loads. These results are used to evaluate the response of the test cell structure for the postulated over pressurization of the test-cell due to the blast load generated in case of 64 gm hydrogen ignition for which additional sets of computations were

Experimental Setup with Transient Behavior of Fuel Cladding of SFR

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

Science.gov (United States)

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

2017-11-01

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

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)

Ultrafast triggered transient energy storage by atomic layer deposition into porous silicon for integrated transient electronics

Science.gov (United States)

Douglas, Anna; Muralidharan, Nitin; Carter, Rachel; Share, Keith; Pint, Cary L.

2016-03-01

Here we demonstrate the first on-chip silicon-integrated rechargeable transient power source based on atomic layer deposition (ALD) coating of vanadium oxide (VOx) into porous silicon. A stable specific capacitance above 20 F g-1 is achieved until the device is triggered with alkaline solutions. Due to the rational design of the active VOx coating enabled by ALD, transience occurs through a rapid disabling step that occurs within seconds, followed by full dissolution of all active materials within 30 minutes of the initial trigger. This work demonstrates how engineered materials for energy storage can provide a basis for next-generation transient systems and highlights porous silicon as a versatile scaffold to integrate transient energy storage into transient electronics.Here we demonstrate the first on-chip silicon-integrated rechargeable transient power source based on atomic layer deposition (ALD) coating of vanadium oxide (VOx) into porous silicon. A stable specific capacitance above 20 F g-1 is achieved until the device is triggered with alkaline solutions. Due to the rational design of the active VOx coating enabled by ALD, transience occurs through a rapid disabling step that occurs within seconds, followed by full dissolution of all active materials within 30 minutes of the initial trigger. This work demonstrates how engineered materials for energy storage can provide a basis for next-generation transient systems and highlights porous silicon as a versatile scaffold to integrate transient energy storage into transient electronics. Electronic supplementary information (ESI) available: (i) Experimental details for ALD and material fabrication, ellipsometry film thickness, preparation of gel electrolyte and separator, details for electrochemical measurements, HRTEM image of VOx coated porous silicon, Raman spectroscopy for VOx as-deposited as well as annealed in air for 1 hour at 450 °C, SEM and transient behavior dissolution tests of uniformly coated VOx on

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

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the third of the NRC L9 series of experiments on Anticipated Transients with Multiple Failures. Loss-of-feedwater effects were studied. The experiment was conducted on 7 April 1982

LOFT/L6-7, Loss of Fluid Test, Anticipated Transients with Multiple Failures

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the seventh in the NRC L6 Series of Anticipated Transients experiments. Rapid secondary side induced cooldown was studied. The experiment was conducted on 31 September 1981

ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3.06.6B - transient film boiling in upflow

International Nuclear Information System (INIS)

Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

1982-05-01

Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

Thermohydraulic tests in nuclear fuel model

International Nuclear Information System (INIS)

Ladeira, L.C.D.; Navarro, M.A.

1984-01-01

The main experimental works performed in the Thermohydraulics Laboratory of the NUCLEBRAS Nuclear Technology Development Center, in the field of thermofluodynamics are briefly described. These works include the performing of steady-state flow tests in single tube test sections, and the design and construction of a rod bundle test section, which will be also used for those kind of testes. Mention is made of the works to be performed in the near future, related to steady-state and transient flow tests. (Author) [pt

Instrumented Taylor anvil-on-rod impact tests for validating applicability of standard strength models to transient deformation states

Science.gov (United States)

Eakins, D. E.; Thadhani, N. N.

2006-10-01

Instrumented Taylor anvil-on-rod impact tests have been conducted on oxygen-free electronic copper to validate the accuracy of current strength models for predicting transient states during dynamic deformation events. The experiments coupled the use of high-speed digital photography to record the transient deformation states and laser interferometry to monitor the sample back (free surface) velocity as a measure of the elastic/plastic wave propagation through the sample length. Numerical continuum dynamics simulations of the impact and plastic wave propagation employing the Johnson-Cook [Proceedings of the Seventh International Symposium on Ballistics, 1983, The Netherlands (Am. Def. Prep. Assoc. (ADPA)), pp. 541-547], Zerilli-Armstrong [J. Appl. Phys. C1, 1816 (1987)], and Steinberg-Guinan [J. Appl. Phys. 51, 1498 (1980)] constitutive equations were used to generate transient deformation profiles and the free surface velocity traces. While these simulations showed good correlation with the measured free surface velocity traces and the final deformed sample shape, varying degrees of deviations were observed between the photographed and calculated specimen profiles at intermediate deformation states. The results illustrate the usefulness of the instrumented Taylor anvil-on-rod impact technique for validating constitutive equations that can describe the path-dependent deformation response and can therefore predict the transient and final deformation states.

LWR fuel performance during anticipated transients with scram

International Nuclear Information System (INIS)

Martinson, Z.R.; McCardell, R.K.; MacDonanl, P.E.; Rowland, T.C.; Tokar, M.

1983-01-01

Operational transients occur occasionally in light water reactors when minor malfunctions of certain system components affect the reactor core. Potential effects of such malfunctions include a loss of the secondary heat sink, an increase in system pressure, and, in boiling water reactors, void collapse and a brief increase in reactor power. The most severe postulated Boiling Water Reactor (BWR) anticipated transient is characterized by a power peak of up to 495% rated power for about 1 second (according to a recent General Electric Co., generic analysis). The results of a series of fuel behaviour tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory are presented in this paper. Four progressively higher and broader power transients at a constant coolant flow rate were performed. The first transient simulated a BWR-5 turbine trip without steam bypass with fuel rods operating at BWR-6 core average rod powers. The second transient simulated a generator load rejection without steam bypass with fuel rods operating at above core average powers. The last two transients were performed at higher powers than safety analysis predicts to be possible in commercial reactors to be defined failure threshold margins. The test rods did not fail and were not damaged during any of the four transients. (author)

A data processing program for transient sodium boiling and fuel failure propagation tests, (2)

International Nuclear Information System (INIS)

Hasebe, Takeshi; Isozaki, Tadashi; Satoh, Akihiro; Yamaguchi, Katsuhisa; Haga, Kazuo.

1983-01-01

Transient Sodium Boiling Tests and Fuel Failure Propagation Tests are being conducted with the out-of-pile test facility, SIENA, in the Core Safety Section of O-arai Engineering Center. The experimental data are recorded using a digital data acquisition system controlled by a HP-1000E computer. The SICILIAN (Speedy Illustration Code for Inspection Line Anomaly) code was developed to obtain quick graphic outputs of data recorded in the magnetic tapes. The program is written in BASIC and Assembler languages and uses a data processing system composed of a desktop computer HP 9845B, a magnetic tape system, a magnetic disc and an eightcolor plotter. The SICILIAN code enables us to get graphic outputs soon after a run. These outputs are very helpful to inspect anomaly in the instrument circuit and to check the experimental conditions of coming runs. (author)

Fission gas behavior during fast thermal transients

International Nuclear Information System (INIS)

Esteves, R.G.

1976-01-01

The behavior of non-equilibrium fission in fuel elements undergoing fast thermal transients is analyzed. To facilitate the analysis, a new variable, the equilibrium variable (EV) is defined. This variable, together with bubble radius, completely specifies a bubble with respect to its size and equilibrium condition. The analysis is coded using a two-variable (radius and EV) multigroup numerical approximation that accepts as input the time-temperature history, the time-fission rate history, and the time-thermal gradient history of the fuel element. Studies were performed to test the code for convergence with respect to the time interval and the number of groups chosen. For a series of transient simulation studies, the measurements obtained at HEDL (microscopic examination of intragranular porosity in oxide fuel transient-tested in TREAT) are used. Two different transient histories were selected; the first, a high-temperature transient (HTT) with a peak at 2477 0 K and the second, a low-temperature transient (LTT) with a peak-temperature at 2000 0 K. The LTT was simulated for three different conditions: Bubbles were allowed to move via (a) only biased migration, (b) via random migration, and (c) via both mechanisms. The HTT was also run for both mechanisms. The agreement with HEDL microscopic observations was fair for bubbles smaller than 964 A in diameter, and poor for larger bubbles. Bubbles that grew during the heat-up part of the transient were frozen at a larger size during the cool down

Damage behavior of REE-doped W-based material exposed to high-flux transient heat loads

International Nuclear Information System (INIS)

Shi, Jing; Luo, Lai–Ma; Lin, Jin–shan; Zan, Xiang; Zhu, Xiao–yong; Xu, Qiu; Wu, Yu–Cheng

2016-01-01

Pure W and W-Lu alloys were prepared by mechanical alloying (MA) and spark plasma sintering (SPS) technology. The performance and relevant damage mechanism of W-(0%, 2%, 5%, 10%) Lu alloys under transient heat loads were investigated using a laser beam heat load test to simulate the transient events in future nuclear fusion reactors. Scanning electron microscopy was used to observe the morphologies of the damaged surfaces and energy dispersive X-ray spectroscopy was used to conduct composition analysis. Damages to the surface such as cracks, pits, melting layers, Lu-rich droplets, and thermal ablation were observed. A mass of dense fuzz-like nanoparticles formed on the outer region of the laser-exposed area. Recrystallization, grain growth, increased surface roughness, and material erosion were also observed. W-Lu samples with low Lu content demonstrated better thermal performance than pure W, and the degree of damage significantly deteriorated under repetitive transient heat loads.

Compilation of Quality Assurance Documentation for Analyses Performed for the Resumption of Transient Testing Environmental Assessment

Energy Technology Data Exchange (ETDEWEB)

Schafer, Annette L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sondrup, A. Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-11-01

This is a companion document to the analyses performed in support of the environmental assessment for the Resumption of Transient Fuels and Materials Testing. It is provided to allow transparency of the supporting calculations. It provides computer code input and output. The basis for the calculations is documented separately in INL (2013) and is referenced, as appropriate. Spreadsheets used to manipulate the code output are not provided.

Instrument response during overpower transients at TREAT

International Nuclear Information System (INIS)

Meek, C.C.; Bauer, T.H.; Hill, D.J.; Froehle, P.H.; Klickman, A.E.; Tylka, J.P.; Doerner, R.C.; Wright, A.E.

1982-01-01

A program to empirically analyze data residuals or noise to determine instrument response that occurs during in-pile transient tests is out-lined. As an example, thermocouple response in the Mark III loop during a severe overpower transient in TREAT is studied both in frequency space and in real-time. Time intervals studied included both constant power and burst portions of the power transient. Thermocouple time constants were computed. Benefits and limitations of the method are discussed

Congestion management enhancing transient stability of power systems

International Nuclear Information System (INIS)

Esmaili, Masoud; Shayanfar, Heidar Ali; Amjady, Nima

2010-01-01

In a competitive electricity market, where market parties try to maximize their profits, it is necessary to keep an acceptable level of power system security to retain the continuity of electricity services to customers at a reasonable cost. Congestion in a power system is turned up due to network limits. After relieving congestion, the network may be operated with a reduced transient stability margin because of increasing the contribution of risky participants. In this paper, a novel congestion management method based on a new transient stability criterion is introduced. Using the sensitivity of corrected transient stability margin with respect to generations and demands, the proposed method so alleviates the congestion that the network can more retain its transient security compared with earlier methods. The proposed transient stability index is constructed considering the likelihood of credible faults. Indeed, market parties participate by their security-effective bids rather than raw bids. Results of testing the proposed method along with the earlier ones on the New-England test system elaborate the efficiency of the proposed method from the viewpoint of providing a better transient stability margin with a lower security cost. (author)

Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

Energy Technology Data Exchange (ETDEWEB)

Ritter, S.; Seifert, H. P

2002-03-01

Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na{sub 2}SO{sub 4} transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm, SO{sub 4}{sup 2-} < 0.6 ppb). Then a sulphate transient was applied. The duration ({approx} 300 h) and the amount of sulphate (SO{sub 4}{sup 2-} = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 {mu}S/cm to less than 0.15 {mu}S/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the

Transient Micromotors That Disappear When No Longer Needed.

Science.gov (United States)

Chen, Chuanrui; Karshalev, Emil; Li, Jinxing; Soto, Fernando; Castillo, Roxanne; Campos, Isaac; Mou, Fangzhi; Guan, Jianguo; Wang, Joseph

2016-11-22

Transient self-destroyed micromotors that autonomously disappear in biological media at controlled rates upon completing their task, without leaving a toxic residue, are presented. The propulsion and degradation characteristics of the self-destroyed Mg/ZnO, Mg/Si, and Zn/Fe Janus micromotors and single-component Zn micromotors are described. The degradation of the Janus micromotors relies on the different corrosion rates of their core-shell components. Inductively coupled plasma optical emission spectrometry measurements are used to probe the time-dependent degradation of the different constituents of the micromotors. The toxicity of the transient micromotors is discussed toward their potential use in biomedical applications. This concept of transient micromotors offers considerable potential for diverse practical applications in the near future.

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

International Nuclear Information System (INIS)

Wambach, J.L.

1980-01-01

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

Transient left ventricular apical ballooning and exercise induced hypertension during treadmill exercise testing: is there a common hypersympathetic mechanism?

Directory of Open Access Journals (Sweden)

Oh Jae K

2008-07-01

Full Text Available Abstract Objective To describe two cases of Takotsubo like myocardial contractile pattern during exercise stress test secondary to hypertensive response. Background Treadmill exercise testing is known to cause sympathetic stimulation, leading to increased levels of catecholamine, resulting in alteration in vascular tone. Hypertensive response during exercise testing can cause abnormal consequences, resulting in false positive results. Cases We present the cases of two patients experiencing apical and basal akinesis during exercise stress echocardiography, in whom normal wall motion response was observed on subsequent pharmacologic stress testing. The first patient developed transient left ventricular (LV apical akinesis during exercise stress echocardiography. Due to high suspicion that this abnormality might be secondary to hypertensive response, pharmacologic stress testing was performed after three days, which was completely normal and showed no such wall motion abnormality. Qualitative assessment of myocardial perfusion using contrast was also performed, which showed good myocardial blood flow, indicating low probability for significant obstructive coronary artery disease. The second patient developed LV basal akinesis as a result of hypertensive response during exercise testing. Coronary angiogram was not performed in either patient due to low suspicion for coronary artery disease, and subsequently negative stress studies. Results Transient stress induced cardiomyopathy can develop secondary to hypertensive response during exercise stress testing. Conclusion These cases provide supporting evidence to the hyper-sympathetic theory of left ventricular ballooning syndrome.

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

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

Behavior of mixed-oxide fuel elements during an overpower transient

International Nuclear Information System (INIS)

Tsai, H.; Shikakura, S.

1993-01-01

A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

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

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.)

Turbofan compressor dynamics during afterburner transients

Science.gov (United States)

Kurkov, A. P.

1976-01-01

The effects of afterburner light-off and shut-down transients on the compressor stability are investigated. The reported experimental results are based on detailed high response pressure and temperature measurements on the TF30-P-3 turbofan engine. The tests were performed in an altitude test chamber simulating high altitude engine operation. It is shown that during both types of transients, flow breaks down in the forward part of the fan bypass duct. At a sufficiently low engine inlet pressure this resulted in a compressor stall. Complete flow breakdown within the compressor was preceded by a rotating stall. At some locations in the compressor, rotating stall cells initially extended only through part of the blade span. For the shutdown transient the time between first and last detected occurrence of rotating stall is related to the flow Reynolds number. An attempt was made to deduce the number and speed of propagation of rotating stall cells.

Test facilities for future linear colliders

International Nuclear Information System (INIS)

Ruth, R.D.

1995-12-01

During the past several years there has been a tremendous amount of progress on Linear Collider technology world wide. This research has led to the construction of the test facilities described in this report. Some of the facilities will be complete as early as the end of 1996, while others will be finishing up around the end 1997. Even now there are extensive tests ongoing for the enabling technologies for all of the test facilities. At the same time the Linear Collider designs are quite mature now and the SLC is providing the key experience base that can only come from a working collider. All this taken together indicates that the technology and accelerator physics will be ready for a future Linear Collider project to begin in the last half of the 1990s

Cirrhosis Diagnosis and Liver Fibrosis Staging: Transient Elastometry Versus Cirrhosis Blood Test.

Science.gov (United States)

Calès, Paul; Boursier, Jérôme; Oberti, Frédéric; Bardou, Derek; Zarski, Jean-Pierre; de Lédinghen, Victor

2015-07-01

Elastometry is more accurate than blood tests for cirrhosis diagnosis. However, blood tests were developed for significant fibrosis, with the exception of CirrhoMeter developed for cirrhosis. We compared the performance of Fibroscan and CirrhoMeter, and classic binary cirrhosis diagnosis versus new fibrosis staging for cirrhosis diagnosis. The diagnostic population included 679 patients with hepatitis C and liver biopsy (Metavir staging and morphometry), Fibroscan, and CirrhoMeter. The prognostic population included 1110 patients with chronic liver disease and both tests. Binary diagnosis: AUROCs for cirrhosis were: Fibroscan: 0.905; CirrhoMeter: 0.857; and P=0.041. Accuracy (Youden cutoff) was: Fibroscan: 85.4%; CirrhoMeter: 79.2%; and PFibrosis classification provided 6 classes (F0/1, F1/2, F2±1, F3±1, F3/4, and F4). Accuracy was: Fibroscan: 88.2%; CirrhoMeter: 88.8%; and P=0.77. A simplified fibrosis classification comprised 3 categories: discrete (F1±1), moderate (F2±1), and severe (F3/4) fibrosis. Using this simplified classification, CirrhoMeter predicted survival better than Fibroscan (respectively, χ=37.9 and 19.7 by log-rank test), but both predicted it well (Ptest). Comparison: binary diagnosis versus fibrosis classification, respectively, overall accuracy: CirrhoMeter: 79.2% versus 88.8% (PFibrosis classification should be preferred over binary diagnosis. A cirrhosis-specific blood test markedly attenuates the accuracy deficit for cirrhosis diagnosis of usual blood tests versus transient elastometry, and may offer better prognostication.

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)

Dynamic behaviour of mono bucket foundations subjected to combined transient loading

DEFF Research Database (Denmark)

Nielsen, Søren Dam; Ibsen, Lars Bo; Nielsen, Benjaminn Nordahl

2015-01-01

This article presents the results from small scale testing, investigating the effect of transient combined loading of a bucketfoundation. The tests are performed inside a pressure tank at Aalborg University, Denmark. The bucket foundation was installed in dense water saturated sand and transient ...

Radiation-Induced Transient Effects in Near Infrared Focal Plane Arrays

Science.gov (United States)

Reed, Robert A.; Pickel, J.; Marshall, P.; Waczynski, A.; McMurray, R.; Gee, G.; Polidan, E.; Johnson, S.; McKeivey, M.; Ennico, K.;

Real time simulator for material testing reactor

Energy Technology Data Exchange (ETDEWEB)

Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

2012-03-15

Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

Real time simulator for material testing reactor

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo

2012-01-01

Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

Evaluating Past and Future USCG Use of Ohmsett Test Facility

Science.gov (United States)

2016-10-01

of Pages 22 22. Price Evaluating Past and Future USCG Use of Ohmsett Test Facility iv UNCLAS//Public | | CG-926 RDC | M. Fitzpatrick, et al...Opportunity Skimming System WEC Wave energy converter Evaluating Past and Future USCG Use of Ohmsett Test Facility x UNCLAS//Public | | CG-926 RDC | M...Date Summary of Effort OCT-NOV 1993 Vessel of Opportunity Skimming System (VOSS) (5 Weeks) APR-JUN 1996 Spilled Oil Recovery System (SORS) (8 Weeks

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

Automated transient identification in the Dark Energy Survey

Energy Technology Data Exchange (ETDEWEB)

Goldstein, D. A. [Univ. of California, Berkeley, CA (United States); Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). et al.

2015-08-20

We describe an algorithm for identifying point-source transients and moving objects on reference-subtracted optical images containing artifacts of processing and instrumentation. The algorithm makes use of the supervised machine learning technique known as Random Forest. We present results from its use in the Dark Energy Survey Supernova program (DES-SN), where it was trained using a sample of 898,963 signal and background events generated by the transient detection pipeline. After reprocessing the data collected during the first DES-SN observing season (2013 September through 2014 February) using the algorithm, the number of transient candidates eligible for human scanning decreased by a factor of 13.4, while only 1.0 percent of the artificial Type Ia supernovae (SNe) injected into search images to monitor survey efficiency were lost, most of which were very faint events. Here we characterize the algorithm's performance in detail, and we discuss how it can inform pipeline design decisions for future time-domain imaging surveys, such as the Large Synoptic Survey Telescope and the Zwicky Transient Facility.

Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options

International Nuclear Information System (INIS)

Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J.; Parma, Edward J.Jr; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

2007-01-01

The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents

Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.

Energy Technology Data Exchange (ETDEWEB)

Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J. (Idaho National Laboratory, Idaho Falls, ID); Parma, Edward J., Jr.; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

2007-10-01

The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.

Cultural Resource Investigations for the Resumption of Transient Testing of Nuclear Fuels and Material at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Pace, Brenda R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williams, Julie B. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-11-01

The U. S. Department of Energy (DOE) has a need to test nuclear fuels under conditions that subject them to short bursts of intense, high-power radiation called ‘transient testing’ in order to gain important information necessary for licensing new nuclear fuels for use in U.S. nuclear power plants, for developing information to help improve current nuclear power plant performance and sustainability, for improving the affordability of new generation reactors, for developing recyclable nuclear fuels, and for developing fuels that inhibit any repurposing into nuclear weapons. To meet this mission need, DOE is considering alternatives for re-use and modification of existing nuclear reactor facilities to support a renewed transient testing program. One alternative under consideration involves restarting the Transient Reactor Test (TREAT) reactor located at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL) site in southeastern Idaho. This report summarizes cultural resource investigations conducted by the INL Cultural Resource Management Office in 2013 to support environmental review of activities associated with restarting the TREAT reactor at the INL. These investigations were completed in order to identify and assess the significance of cultural resources within areas of potential effect associated with the proposed action and determine if the TREAT alternative would affect significant cultural resources or historic properties that are eligible for nomination to the National Register of Historic Places. No archaeological resources were identified in the direct area of potential effects for the project, but four of the buildings proposed for modifications are evaluated as historic properties, potentially eligible for nomination to the National Register of Historic Places. This includes the TREAT reactor (building #), control building (building #), guardhouse (building #), and warehouse (building #). The proposed re-use of these historic

Impacts of different rainfall patterns on hyporheic zone under transient conditions

Science.gov (United States)

Liu, Suning; Chui, Ting Fong May

2018-06-01

The hyporheic zone (HZ) plays an important role in stream ecology. Previous studies have mainly focused on the factors influencing the HZ in the steady state. However, the exchange between surface water and groundwater in the HZ can become transient during a storm. This study investigates the impacts of different rainfall patterns (varying in intensity and duration) on the HZ under transient conditions. A two-dimensional numerical model of a 10-m long and 2-m deep domain is developed, in which the streambed consists of a series of dunes. Brinkman-Darcy and Navier-Stokes equations are respectively solved for groundwater and surface water, and velocity and pressure are coupled at the interface (i.e., the streambed surface). To compare the results under different transient conditions, this study proposes two indicators, i.e., the influential time (IT, the time required for the HZ to return to its initial state once it starts to change) and the influential depth (ID, the maximum increment in the HZ depth). To detect the extent to which the HZ undergoes significant spatial changes, moving split-window and inflection point tests are conducted. The results indicate that rainfall intensity (RI) and rainfall duration (RD) both display logarithmic relationships with the IT and ID with high coefficients of determination, but only between certain lower and upper thresholds of the RI and RD. Moreover, the distributions of the IT and ID as a function of the RI and RD are mapped using the surface spline and kriging interpolation methods to facilitate future prediction of the IT and ID. In addition, it is observed that the IT has a linear negative correlation with the groundwater response while the ID is not affected by different groundwater responses. All of the derived relationships can be used to predict the impacts of a future rainfall event on the HZ.

TRAC analyses of severe overcooling transients for the Oconee-1 PWR

Energy Technology Data Exchange (ETDEWEB)

Ireland, J R [comp.

1985-05-01

This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.

TRAC analyses of severe overcooling transients for the Oconee-1 PWR

International Nuclear Information System (INIS)

Ireland, J.R.

1985-05-01

This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs

Boiling and burnout phenomena under transient heat input, 1

International Nuclear Information System (INIS)

Aoki, Shigebumi; Kozawa, Yoshiyuki; Iwasaki, Hideaki.

1976-01-01

In order to simulate the thermo-hydrodynamic conditions at reactor power excursions, a test piece was placed in a forced convective channel and heated with exponential power inputs. The boiling heat transfer and the burnout heat flux under the transient heat input were measured, and pressure and water temperature changes in the test section were recorded at the same time. Following experimental results were obtained; (1) Transient boiling heat transfer characteristics at high heat flux stayed on the stationary nucleate boiling curve of each flow condition, or extrapolated line of the curves. (2) Transient burnout heat flux increased remarkably with decreasing heating-time-constant, when the flow rate was lower and the subcooling was higher. (3) Transient burnout phenomena were expressed with the relation of (q sub(max) - q sub(sBO)) tau = constant at several flow conditions. This relation was derived from the stationary burnout mechanism of pool boiling. (auth.)

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

Analysis of the FFTF primary pipe rupture transients

International Nuclear Information System (INIS)

Perkins, K.R.; Bari, R.A.; Chen, L.C.; Albright, D.C.

1979-01-01

The response of the Fast Flux Test Facility (FFTF) to hypothetical ruptures of the high pressure primary piping has been analyzed using two LMFBR plant systems codes, namely IANUS and DEMO. Comparisons of the average channel temperatures predicted by the two codes show good agreement for identical transients. However, the hot channel temperatures predicted by DEMO are about 60K higher than the corresponding IANUS predictions for severe transients. This difference is attributed to the dynamic hot channel factors employed in DEMO which discount the thermal inertia of the duct walls for rapid transients. DEMO also predicts more severe transients for hot-leg ruptures in FFTF than previously reported analyses for the CRBR

A new transiently chaotic flow with ellipsoid equilibria

Science.gov (United States)

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

2018-03-01

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

RELAP5 capabilities in thermal-hydraulic prediction of SBWR containment behaviour: PANDA steady state and transient tests evaluation

International Nuclear Information System (INIS)

Faluomi, V.; Aksan, S.N.

2000-01-01

This paper summarizes the results of the qualification activity of RELAP5/Mod3.2 code performed using PANDA steady state and integral test experimental data. The steady state tests evaluate the PCC performances in removing decay heat power in presence and in absence of non-condensable gases, while the considered integral test (M3) simulates the transient following a break in the main steam line of the SBWR, using, as nominal initial conditions, those calculated for the SBWR under SSAR assumptions at one hour into the LOCA. The results obtained simulating both types of tests show a rather good and robust overall code behavior both in the simulation of steady state test and in the representation of the integral test considered: most of the main experimental results (WW/DW pressures, PCC heat exchange) were well represented by the code. The different studies performed indicated that: Different models of PCC pool lead a different trend of system pressure, and sometimes to an opening of vacuum breaker valves, that does not occur in the transient; The code underestimate the heat exchanged between PCC pool and tubes: n the considered test the system pressure is slightly overestimated (maximum 2% more than the experimental value). This fact is also proved by the differences in the temperature of the condensing mixture in the PCC, quite large in all the performed studies; The treatment of the non condensable gases, as implemented in the code, lead some errors in the calculation of the heat transfer coefficient in the PCC components and generally slow down the overall calculation. In general terms, the RELAP5/Mod3.2 was found to be suitable to represent the SBWR containment behavior under the conditions specified in the experimental side. (author)

Model for transient creep of southeastern New Mexico rock salt

International Nuclear Information System (INIS)

Herrmann, W.; Wawersik, W.R.; Lauson, H.S.

1980-11-01

In a previous analysis, existing experimental data pertaining to creep tests on rock salt from the Salado formation of S.E. New Mexico were fitted to an exponential transient creep law. While very early time portions of creep strain histories were not fitted very well for tests at low temperatures and stresses, initial creep rates in particular generally being underestimated, the exponential creep law has the property that the transient creep strain approaches a finite limit with time, and is therefore desirable from a creep modelling point of view. In this report, an analysis of transient creep is made. It is found that exponential transient creep can be related to steady-state creep through a universal creep curve. The resultant description is convenient for creep analyses where very early time behavior is not important

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

Transient Hypothyroidism after Radioiodine for Graves' Disease: Challenges in Interpreting Thyroid Function Tests.

Science.gov (United States)

Sheehan, Michael T; Doi, Suhail A R

2016-03-01

Graves' disease is the most common cause of hyperthyroidism and is often managed with radioactive iodine (RAI) therapy. With current dosing schemes, the vast majority of patients develop permanent post-RAI hypothyroidism and are placed on life-long levothyroxine therapy. This hypothyroidism typically occurs within the first 3 to 6 months after RAI therapy is administered. Indeed, patients are typically told to expect life-long thyroid hormone replacement therapy to be required within this timeframe and many providers expect this post-RAI hypothyroidism to be complete and permanent. There is, however, a small subset of patients in whom a transient post-RAI hypothyroidism develops which, initially, presents exactly as the typical permanent hypothyroidism. In some cases the transient hypothyroidism leads to a period of euthyroidism of variable duration eventually progressing to permanent hypothyroidism. In others, persistent hyperthyroidism requires a second dose of RAI. Failure to appreciate and recognize the possibility of transient post-RAI hypothyroidism can delay optimal and appropriate treatment of the patient. We herein describe five cases of transient post-RAI hypothyroidism which highlight this unusual sequence of events. Increased awareness of this possible outcome after RAI for Graves' disease will help in the timely management of patients. © 2016 Marshfield Clinic.

Transient Hypothyroidism after Radioiodine for Graves’ Disease: Challenges in Interpreting Thyroid Function Tests

Science.gov (United States)

Sheehan, Michael T.; Doi, Suhail A.R.

2016-01-01

Graves’ disease is the most common cause of hyperthyroidism and is often managed with radioactive iodine (RAI) therapy. With current dosing schemes, the vast majority of patients develop permanent post-RAI hypothyroidism and are placed on life-long levothyroxine therapy. This hypothyroidism typically occurs within the first 3 to 6 months after RAI therapy is administered. Indeed, patients are typically told to expect life-long thyroid hormone replacement therapy to be required within this timeframe and many providers expect this post-RAI hypothyroidism to be complete and permanent. There is, however, a small subset of patients in whom a transient post-RAI hypothyroidism develops which, initially, presents exactly as the typical permanent hypothyroidism. In some cases the transient hypothyroidism leads to a period of euthyroidism of variable duration eventually progressing to permanent hypothyroidism. In others, persistent hyperthyroidism requires a second dose of RAI. Failure to appreciate and recognize the possibility of transient post-RAI hypothyroidism can delay optimal and appropriate treatment of the patient. We herein describe five cases of transient post-RAI hypothyroidism which highlight this unusual sequence of events. Increased awareness of this possible outcome after RAI for Graves’ disease will help in the timely management of patients. PMID:26864507

Standard Test Method for Measuring Extreme Heat-Transfer Rates from High-Energy Environments Using a Transient, Null-Point Calorimeter

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers the measurement of the heat-transfer rate or the heat flux to the surface of a solid body (test sample) using the measured transient temperature rise of a thermocouple located at the null point of a calorimeter that is installed in the body and is configured to simulate a semi-infinite solid. By definition the null point is a unique position on the axial centerline of a disturbed body which experiences the same transient temperature history as that on the surface of a solid body in the absence of the physical disturbance (hole) for the same heat-flux input. 1.2 Null-point calorimeters have been used to measure high convective or radiant heat-transfer rates to bodies immersed in both flowing and static environments of air, nitrogen, carbon dioxide, helium, hydrogen, and mixtures of these and other gases. Flow velocities have ranged from zero (static) through subsonic to hypersonic, total flow enthalpies from 1.16 to greater than 4.65 × 101 MJ/kg (5 × 102 to greater than 2 × 104 ...

Transient Ischemic Attack Caused by Iron Deficiency Anemia

Directory of Open Access Journals (Sweden)

Ufuk Emre

2006-02-01

Full Text Available Transient Ischemic Attack Caused by Iron Deficiency Anemia Transient ischemic attacks are episodes of transient focal ischemia involving the brain or brainstem. They are commonly two to thirty minutes in duration and lasting less than 24 hours. Anemia of iron deficiency isn’t frequently cause for transient ischemic attack. It has been reported as a risk factor for childhood ischemic strokes. In the iron deficiency anemia, T‹A may develop as result of hypercoagulable state and increased viscosity that is caused by anemic hypoxia that is result of reduce hemoglobine level, seconder thrombosis and microcytose As iron deficiency anemia has been reported so rarely in adult patients with transient ischemic attacks as a cause, we aimed to discuss the clinical and outcome features of two cases with iron deficiency anemia and transient ischemic attacks in this study. Materials and methods: Routine neurologic examination, biochemical screen, serological tests, vasculitic markers, thyroid function tests, vitamin B 12 level, cranial imaging, vertebral carotid doppler USG examination was conducted in the two patients. Anemia of iron deficiency was found as the only risk factor for TIA and the two patients were treated with replacement of iron and antiagregan therapy. Neurological examination revealed no abnormality through the two years of follow-up. The iron deficiency anemia may be cause of many neurologic problems such a irritability, lethargy, headache, development retardation except from T‹A. In the iron deficiency anemia, early diagnosis and treatment is important

Further elucidation of nanofluid thermal conductivity measurement using a transient hot-wire method apparatus

Science.gov (United States)

Yoo, Donghoon; Lee, Joohyun; Lee, Byeongchan; Kwon, Suyong; Koo, Junemo

2018-02-01

The Transient Hot-Wire Method (THWM) was developed to measure the absolute thermal conductivity of gases, liquids, melts, and solids with low uncertainty. The majority of nanofluid researchers used THWM to measure the thermal conductivity of test fluids. Several reasons have been suggested for the discrepancies in these types of measurements, including nanofluid generation, nanofluid stability, and measurement challenges. The details of the transient hot-wire method such as the test cell size, the temperature coefficient of resistance (TCR) and the sampling number are further investigated to improve the accuracy and consistency of the measurements of different researchers. It was observed that smaller test apparatuses were better because they can delay the onset of natural convection. TCR values of a coated platinum wire were measured and statistically analyzed to reduce the uncertainty in thermal conductivity measurements. For validation, ethylene glycol (EG) and water thermal conductivity were measured and analyzed in the temperature range between 280 and 310 K. Furthermore, a detailed statistical analysis was conducted for such measurements, and the results confirmed the minimum number of samples required to achieve the desired resolution and precision of the measurements. It is further proposed that researchers fully report the information related to their measurements to validate the measurements and to avoid future inconsistent nanofluid data.

Aeroelastic Modeling of a Nozzle Startup Transient

Science.gov (United States)

Wang, Ten-See; Zhao, Xiang; Zhang, Sijun; Chen, Yen-Sen

2014-01-01

Lateral nozzle forces are known to cause severe structural damage to any new rocket engine in development during test. While three-dimensional, transient, turbulent, chemically reacting computational fluid dynamics methodology has been demonstrated to capture major side load physics with rigid nozzles, hot-fire tests often show nozzle structure deformation during major side load events, leading to structural damages if structural strengthening measures were not taken. The modeling picture is incomplete without the capability to address the two-way responses between the structure and fluid. The objective of this study is to develop a tightly coupled aeroelastic modeling algorithm by implementing the necessary structural dynamics component into an anchored computational fluid dynamics methodology. The computational fluid dynamics component is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, while the computational structural dynamics component is developed under the framework of modal analysis. Transient aeroelastic nozzle startup analyses at sea level were performed, and the computed transient nozzle fluid-structure interaction physics presented,

Environmental, Transient, Three-Dimensional, Hydrothermal, Mass Transport Code - FLESCOT

Energy Technology Data Exchange (ETDEWEB)

Onishi, Yasuo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Glass, Kevin A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Eyler, L. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Okumura, Masahiko [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-03-28

The purpose of the project was to modify and apply the transient, three-dimensional FLESCOT code to be able to effectively simulate cesium behavior in Fukushima lakes/dam reservoirs, river mouths, and coastal areas. The ultimate objective of the FLESCOT simulation is to predict future changes of cesium accumulation in Fukushima area reservoirs and costal water. These evaluation results will assist ongoing and future environmental remediation activities and policies in a systematic and comprehensive manner.

Change in geometrical parameters of WWER high burnup fuel rods under operational conditions and transient testing

International Nuclear Information System (INIS)

Kanashov, B.; Amosov, S.; Lyadov, G.; Markov, D.; Ovchinnikov, V; Polenok, V.; Smirnov, A.; Sukhikh, A.; Bek, E.; Yenin, A.; Novikov, V.

2001-01-01

The paper discusses changes in fuel rods geometric parameters as result of operation conditions and burnups. The degree of geometry variability of fuel rods, cladding and column is one of the most important characteristics affecting fuel serviceability. On the other hand, changes in fuel rod geometric parameters influence fuel temperature, fission gas release, fuel-to-cladding stress strained state as well as the degree of interaction with FA skeleton elements and skeleton rigidity. Change in fuel-to-cladding gap is measured using compression technique. The axial distribution of fuel-to-cladding gap demonstrates the largest decrease of the gap in the region 500 to 2000 mm from the bottom of the fuel rod (WWER-440) and in the region of 500 to 3000 mm for WWER-1000. The cladding material creep in WWER fuel rods together with the radiation growth results in fuel rod cladding elongation. A set of transient tests for spent WWER-440 and WWER-1000 fuel rods carried out in SSC RIAR during a period 1995-1999, with the aim to estimate the changes in geometric parameters of FRs. The estimation of changes in outer diameter of cladding and fuel column and fuel-to-cladding gap are performed in transient conditions (changes in linear power range of 180 to 400 W/cm) for both WWER-440 and WWER-1000. WWER-440 fuel rods having the same burnup and close fuel-cladding contact before testing are subjected to considerable hoop cladding strain in testing up to 300 W/cm. But the hoop strain does not grow due to the structural changes in fuel column and decrease in central hole diameter occurred when the power is higher

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.)

Test of the Weak Equivalence Principle using LIGO observations of GW150914 and Fermi observations of GBM transient 150914

Directory of Open Access Journals (Sweden)

Molin Liu

2017-07-01

Full Text Available About 0.4 s after the Laser Interferometer Gravitational-Wave Observatory (LIGO detected a transient gravitational-wave (GW signal GW150914, the Fermi Gamma-ray Burst Monitor (GBM also found a weak electromagnetic transient (GBM transient 150914. Time and location coincidences favor a possible association between GW150904 and GBM transient 150914. Under this possible association, we adopt Fermi's electromagnetic (EM localization and derive constraints on possible violations of the Weak Equivalence Principle (WEP from the observations of two events. Our calculations are based on four comparisons: (1 The first is the comparison of the initial GWs detected at the two LIGO sites. From the different polarizations of these initial GWs, we obtain a limit on any difference in the parametrized post-Newtonian (PPN parameter Δγ≲10−10. (2 The second is a comparison of GWs and possible EM waves. Using a traditional super-Eddington accretion model for GBM transient 150914, we again obtain an upper limit Δγ≲10−10. Compared with previous results for photons and neutrinos, our limits are five orders of magnitude stronger than those from PeV neutrinos in blazar flares, and seven orders stronger than those from MeV neutrinos in SN1987A. (3 The third is a comparison of GWs with different frequencies in the range [35 Hz, 250 Hz]. (4 The fourth is a comparison of EM waves with different energies in the range [1 keV, 10 MeV]. These last two comparisons lead to an even stronger limit, Δγ≲10−8. Our results highlight the potential of multi-messenger signals exploiting different emission channels to strengthen existing tests of the WEP.

Transient thermal creep of nuclear reactor pressure vessel type concretes

International Nuclear Information System (INIS)

Khoury, G.A.

1983-01-01

The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and

Testing the utility of fluorescent proteins in Mimulus lewisii by an Agrobacterium-mediated transient assay.

Science.gov (United States)

Ding, Baoqing; Yuan, Yao-Wu

2016-04-01

The Agrobacterium -mediated transient expression assay by leaf infiltration in Mimulus lewisii is robust. Fluorescent proteins EGFP, EYFP and DsRed give bright fluorescence signals in the infiltrated tissue. Mimulus lewisii is an emerging developmental genetic model system. Recently developed genomic and genetic resources and a stable transformation protocol have greatly facilitated the identification and functional characterization of genes controlling the development of ecologically important floral traits using this species. To further expedite gene and protein function analyses in M. lewisii, we adopted and simplified the Agrobacterium-mediated transient gene expression method routinely used in tobacco plants. With the validated transient assay, we examined the performance of fluorescent proteins EGFP, EYFP and DsRed in M. lewisii. All three proteins gave bright fluorescence signals when transiently expressed in agroinfiltrated leaves. Furthermore, we demonstrated the utility of fluorescent proteins in M. lewisii by showing the nuclear localization of Reduced Carotenoid Pigmentation 1 (RCP1), a recently discovered R2R3-MYB transcription factor that regulates carotenoid pigmentation during flower development. Both the transient assay and the fluorescent proteins are valuable additions to the M. lewisii toolbox, making this emerging genetic and developmental model system even more powerful.

The transient radiation effects and hardness of programmed device

International Nuclear Information System (INIS)

Du Chuanhua; Xu Xianguo; Zhao Hailin

2014-01-01

A review and summary of research and development in the investigation of transient ionizing radiation effects in device and cirviut is presented. The transient ionizing radiation effects in two type of programmed device, that's 32 bit Microcontroller and antifuse FPGA, were studied. The expeiment test data indicate: The transient ionizing radiation effects of 32 bit Microcontroller manifested self-motion restart and Latchup, the Latchup threshold was 5 × 10"7 Gy (Si)/s. The transient ionizing radiation effects of FPGA was reset, no Latchup. The relationship of circuit effects to physical mechanisms was analized. A new method of hardness in circiut design was put forward. (authors)

Measuring the permeability of Eleana argillite from area 17, Nevada Test Site, using the transient method

International Nuclear Information System (INIS)

Lin, W.

1978-01-01

Using the transient method, we determine the permeability of high-quartz Eleana argillite from the Nevada Test Site as a function of effective pressure. By comparing calculated and observed pressure decay in the upstream reservoir, we have determined the permeability of intact and fractured specimens at effective pressures ranging from 1.0 to 24.0 MPa. Over this pressure range, Eleana argillite has a low permeability (10 -16 to 10 -19 cm 2 ) when intact and a higher permeability (10 -12 to 10 -17 cm 2 ) with one induced through-going fracture

THE CAUSALITY TEST BETWEEN THE VARIANCES OF SPOT AND FUTURE MARKET PRICES

Directory of Open Access Journals (Sweden)

EMRAH İSMAİL ÇEVİK

2013-06-01

Full Text Available Volatility in financial markets urges importance of risk management with respect to investors and especially firms. Information and interaction between spot and futures markets plays an important role on formation of market prices. In this study, causality and information flows are examined on spot and futures prices of ISE 100 Index, US Dollar, and Euro which are traded at Turkish Derivatives Exchange (VOB. Dynamic causality test that is originally created by Cheung and Ng (1996 is applied. Dynamic causality test results show that in the ISE 100 Index model spot prices affect futures prices and in the exchange model futures prices affect spot prices.

An analysis of feature relevance in the classification of astronomical transients with machine learning methods

Science.gov (United States)

D'Isanto, A.; Cavuoti, S.; Brescia, M.; Donalek, C.; Longo, G.; Riccio, G.; Djorgovski, S. G.

2016-04-01

The exploitation of present and future synoptic (multiband and multi-epoch) surveys requires an extensive use of automatic methods for data processing and data interpretation. In this work, using data extracted from the Catalina Real Time Transient Survey (CRTS), we investigate the classification performance of some well tested methods: Random Forest, MultiLayer Perceptron with Quasi Newton Algorithm and K-Nearest Neighbours, paying special attention to the feature selection phase. In order to do so, several classification experiments were performed. Namely: identification of cataclysmic variables, separation between galactic and extragalactic objects and identification of supernovae.

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...

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

International Nuclear Information System (INIS)

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

1979-01-01

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

In beam test of a transient magnetic field based g factor setup

International Nuclear Information System (INIS)

Saxena, Mansi; Mandal, S.; Siwal, Davinder; Kumar, Chandan; Goyal, Savi; Garg, Ritika; Khushboo; Rohilla, Aman; Kumar, Naveen; Kumar, S.; Chamoli, S.; Mandal, A.; Kumar, Rajesh; Barua, P.; Kumar, R.; Gujjar, R.; Bala, Indu; Singh, R.P.; Muralithar, S.; Rehman, Anisur; Roy, Minakshi

2012-01-01

Magnetic moments or g factors provide substantial information on the microscopic structure of the nuclei. Since the g factor is very different in sign and magnitude for neutrons and protons, therefore they can serve as a good indicator as to which nucleon contributes most to the wave function of that state. There are different techniques available for measuring g factors depending on the lifetime of the state involved. Using transient field technique we can measure the g factor of nuclei having a lifetime of the order of several hundreds of femto seconds. To measure the g factor of such low lifetime states we have designed and fabricated a setup based on the Transient Field Technique Measurement

Mitigation of thermal transients by tube bundle inlet plenum design

International Nuclear Information System (INIS)

Oras, J.J.; Kasza, K.E.

1984-06-01

A multiphase program aimed at investigating the importance of thermal buoyancy to LMFBR steam-generator and heat-exchanger thermal hydraulics under low-flow transient conditions is being conducted in the Argonne Mixing Components Test Facility (MCTF) on a 60 0 sector shell-side flow model of the Westinghouse straight-tube steam generator being developed under the US/DOE large-component development program. A series of shell-side constant-flow thermal-downramp transient tests have been conducted focusing on the phenomenon of thermal-buoyancy-induced-flow channeling. In addition, it was discovered that a shell-inlet flow-distribution plenum can play a significant role in mitigating the severity of a thermal transient entering a steam generator or heat exchanger

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

New modelling of transient well test and rate decline analysis for a horizontal well in a multiple-zone reservoir

International Nuclear Information System (INIS)

Nie, Ren-Shi; Guo, Jian-Chun; Jia, Yong-Lu; Zhu, Shui-Qiao; Rao, Zheng; Zhang, Chun-Guang

2011-01-01

The no-type curve with negative skin of a horizontal well has been found in the current research. Negative skin is very significant to transient well test and rate decline analysis. This paper first presents the negative skin problem where the type curves with negative skin of a horizontal well are oscillatory. In order to solve the problem, we propose a new model of transient well test and rate decline analysis for a horizontal well in a multiple-zone composite reservoir. A new dimensionless definition of r D is introduced in the dimensionless mathematical modelling under different boundaries. The model is solved using the Laplace transform and separation of variables techniques. In Laplace space, the solutions for both constant rate production and constant wellbore pressure production are expressed in a unified formula. We provide graphs and thorough analysis of the new standard type curves for both well test and rate decline analysis; the characteristics of type curves are the reflections of horizontal well production in a multiple-zone reservoir. An important contribution of our paper is that our model removed the oscillation in type curves and thus solved the negative skin problem. We also show that the characteristics of type curves depend heavily on the properties of different zones, skin factor, well length, formation thickness, etc. Our research can be applied to a real case study

Transient accelerating scalar models with exponential potentials

International Nuclear Information System (INIS)

Cui Wen-Ping; Zhang Yang; Fu Zheng-Wen

2013-01-01

We study a known class of scalar dark energy models in which the potential has an exponential term and the current accelerating era is transient. We find that, although a decelerating era will return in the future, when extrapolating the model back to earlier stages (z ≳ 4), scalar dark energy becomes dominant over matter. So these models do not have the desired tracking behavior, and the predicted transient period of acceleration cannot be adopted into the standard scenario of the Big Bang cosmology. When couplings between the scalar field and matter are introduced, the models still have the same problem; only the time when deceleration returns will be varied. To achieve re-deceleration, one has to turn to alternative models that are consistent with the standard Big Bang scenario.

CANDU fuel behaviour under transient conditions

International Nuclear Information System (INIS)

Segel, A.W.L.

1979-04-01

The Canadian R and D program to understand CANDU fuel behaviour under transient conditions is described. Fuel sheath behaviour studies have led to the development of a model of transient plastic strain in inert gas, which integrates the deformation due to several mechanisms. Verification tests demonstrated that on average the model overpredicts strain by 20%. From oxidation kinetics studies a sheath failure embrittlement criterion based on oxygen distribution has been developed. We have also established a rate equation for high-temperature stress-dependent crack formation due to embrittlement of the sheath by beryllium. An electric, simulated fuel element is being used in laboratory tests to characterize the behaviour of fuel in the horizontal. In-reactor, post-dryout tests have been done for several years. There is an axially-segmented, axisymmetric fuel element model in place and a fully two-dimensional code is under development. Laboratory testing of bundles, in its early stages, deals with the effects of geometric distortion and sheath-to-sheath interaction. In-reactor, post-dryout tests of CANDU fuel bundles with extensive central UO 2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)

Incidental transients problems in reactor. Application examples

International Nuclear Information System (INIS)

Marbach, G.

1988-03-01

The fast neutron reactor fuel element qualification should be made not only for nominal operation but also for incidental and accidental transients. Different studies and tests permit to bring this justification such as simulation in hot laboratory after irradiation of irradiated pins or specific tests interpretation [fr

Initial Implementation of Transient VERA-CS

Energy Technology Data Exchange (ETDEWEB)

Gerlach, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Salko, Robert [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-11-01

In this milestone the capabilities of both CTF and MPACT were extended to perform coupled transient calculations. This required several small changes in MPACT to setup the problems correctly, perform the edits correctly, and call the appropriate CTF interfaces in the right order. For CTF, revisions and corrections to the transient timestepping algorithm were made, as well as the addition of a new interface subroutine to allow MPACT to drive CTF at each timestep. With the modifications completed, the initial coupled capability was demonstrated on some problems used for code verification, a hypothetical small mini-core, and a Watts Bar demonstration problem. For each of these cases the results showed good agreement with the previous MPACT internal TH feedback model that relied on a simplified fuel heat conduction model and simplified coolant treatment. After the pulse the results are notably different as expected, where the effects of convection of heat to the coolant can be observed. Areas for future work were discussed, including assessment and development of the CTF dynamic fuel deformation and gap conductance models, addition of suitable transient boiling and CHF models for the rapid heating and cooling rates seen in RIAs, additional validation and demonstration work, and areas for improvement to the code input and output capabilities.

Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

International Nuclear Information System (INIS)

Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

1986-01-01

Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

Association of transient hyperthyroidism and severity of hyperemesis gravidarum.

Science.gov (United States)

Malek, Nor Zila Hassan; Kalok, Aida; Hanafiah, Zainal Abidin; Shah, Shamsul Azhar; Ismail, Nor Azlin Mohamed

2017-03-23

Background Transient non-immune hyperthyroidism in early pregnancy is hyperthyroidism diagnosed for the first time in early pregnancy, without evidence of thyroid autoimmunity or clinical findings of Grave's disease and resolved spontaneously as the pregnancy progressed. Hyperemesis gravidarum (HG) is the commonest cause with 66%-73% of women with severe HG were found to have elevated thyroid function. Materials and methods We conducted a cross sectional study to determine the prevalence of transient hyperthyroidism in patients with hyperemesis gravidarum and its relation to the severity of nausea and vomiting. Severity of nausea and vomiting in pregnancy was assessed using the modified pregnancy-unique quantification of emesis (PUQE) scoring system. Each patient had urine and blood investigations which also included a full blood count and thyroid and renal function tests. Patients with abnormal thyroid function were retested at 20 weeks of gestation. The patients' demographic data, electrolyte levels, thyroid function and their respective PUQE score were analyzed. Results The prevalence of transient hyperthyroidism in women with hyperemesis gravidarum was 4.8%. Although there was a significant association between the severity of the PUQE score and hypokalemia (p = 0.001), there was no significant association with transient hyperthyroidism in early pregnancy (p = 0.072). Free T4 and TSH values of all women with transient hyperthyroidism were normalized by 20 weeks of gestation. Conclusion Transient hyperthyroidism in pregnancy is not significantly associated with the severity of the PUQE score. Women with transient hyperthyroidism in pregnancy are normally clinically euthyroid, hence a routine thyroid function test is unnecessary unless they exhibit clinical signs or symptoms of hyperthyroidism.

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

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.)

Cladding properties under simulated fuel pin transients

International Nuclear Information System (INIS)

Hunter, C.W.; Johnson, G.D.

1975-01-01

A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

Comparison of nine blood tests and transient elastography for liver fibrosis in chronic hepatitis C: the ANRS HCEP-23 study.

Science.gov (United States)

Zarski, Jean-Pierre; Sturm, Nathalie; Guechot, Jérôme; Paris, Adeline; Zafrani, Elie-Serge; Asselah, Tarik; Boisson, Renée-Claude; Bosson, Jean-Luc; Guyader, Dominique; Renversez, Jean-Charles; Bronowicki, Jean-Pierre; Gelineau, Marie-Christine; Tran, Albert; Trocme, Candice; De Ledinghen, Victor; Lasnier, Elisabeth; Poujol-Robert, Armelle; Ziegler, Frédéric; Bourliere, Marc; Voitot, Hélène; Larrey, Dominique; Rosenthal-Allieri, Maria Alessandra; Fouchard Hubert, Isabelle; Bailly, François; Vaubourdolle, Michel

2012-01-01

Blood tests and transient elastography (Fibroscan™) have been developed as alternatives to liver biopsy. This ANRS HCEP-23 study compared the diagnostic accuracy of nine blood tests and transient elastography (Fibroscan™) to assess liver fibrosis, vs. liver biopsy, in untreated patients with chronic hepatitis C (CHC). This was a multicentre prospective independent study in 19 French University hospitals of consecutive adult patients having simultaneous liver biopsy, biochemical blood tests (performed in a centralized laboratory) and Fibroscan™. Two experienced pathologists independently reviewed the liver biopsies (mean length=25±8.4 mm). Performance was assessed using ROC curves corrected by Obuchowski's method. Fibroscan™ was not interpretable in 113 (22%) patients. In the 382 patients having both blood tests and interpretable Fibroscan™, Fibroscan™ performed similarly to the best blood tests for the diagnosis of significant fibrosis and cirrhosis. Obuchowski's measure showed Fibrometer® (0.86), Fibrotest® (0.84), Hepascore® (0.84), and interpretable Fibroscan™ (0.84) to be the most accurate tests. The combination of Fibrotest®, Fibrometer®, or Hepascore® with Fibroscan™ or Apri increases the percentage of well classified patients from 70-73% to 80-83% for significant fibrosis, but for cirrhosis a combination offers no improvement. For the 436 patients having all the blood tests, AUROC's ranged from 0.82 (Fibrometer®) to 0.75 (Hyaluronate) for significant fibrosis, and from 0.89 (Fibrometer® and Hepascore®) to 0.83 (FIB-4) for cirrhosis. Contrarily to blood tests, performance of Fibroscan™ was reduced due to uninterpretable results. Fibrotest®, interpretable Fibroscan™, Fibrometer®, and Hepascore® perform best and similarly for diagnosis of significant fibrosis and cirrhosis. Copyright © 2011 European Association for the Study of the Liver. Published by Elsevier B.V. All rights reserved.

Development of MCP transient operation strategy for the SMART-P

International Nuclear Information System (INIS)

Yoo, S. E.; Choi, B. S.; Kang, H. O.; Yoon, J. H.; Ji, S. K.

2003-01-01

SMART-P MCP(Main Coolant Pump) transient operation strategies are developed. A Modular Modeling System (MMS) computer code is used for the evaluation of the developed operation strategies. In the SMART-P, normal operating modes are classified into MCP high speed(3600 rpm) mode and MCP low speed mode. Also, natural circulation mode is defined as a performance test case. MCP operation transients occur when changing modes from one to another, and system parameters(core power, system pressure, temperature) are having transients. These transients affect on system performance and, in some cases, limit system operation. In this study, MCP operation strategies are developed and obtained acceptable results

Do Magnetic Fields drive high-energy explosive transients?

OpenAIRE

Mundell, Carole

2017-01-01

I will review the current state-of-the-art in real-time, rapid response optical imaging and polarimetric followup of transient sources such as Gamma Ray Bursts. I will interpret current results within the context of the external shock model and present predictions for future mm- and cm-wave radio observatories. Recent observational results from new radio pilot studies will also be presented.

Do Magnetic Fields Drive High-Energy Explosive Transients?

Science.gov (United States)

Mundell, Carole

2017-10-01

I will review the current state-of-the-art in real-time, rapid response optical imaging and polarimetric followup of transient sources such as Gamma Ray Bursts. I will interpret current results within the context of the external shock model and present predictions for future mm- and cm-wave radio observatories. Recent observational results from new radio pilot studies will also be presented.

Transient performance of S-prism

International Nuclear Information System (INIS)

Dubberley, A.E.; Boardman, C.E.; Gamble, R.E.; Hiu, M.M.; Lipps, A.J.; Wu, T.

2001-01-01

S-PRISM is an advanced Fast Reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test of a single Nuclear Steam Supply System (NSSS) for design certification at minimum cost and risk. Based on the success of the previous DOE sponsored Advanced Liquid Metal Reactor (ALMR) program GE has continued to develop and assess the technical viability and economic potential of an up-rated plant called SuperPRISM (S-PRISM). This paper presents the results of transient analyses performed to assess the ability of S-PRISM to accommodate severe accident initiator events. A unique safety capability of S-PRISM is accommodation of the ''higher probability'' accident initiators that led to core melt accidents in prior large LMRs. These events, the Anticipated Transients Without Scram (ATWS) events, are thus the focus of passive safety confirmation analyses. The events included in this assessment are: Unprotected Loss of Flow, Unprotected Loss of Heat Sink, Unprotected Loss of Flow and Heat sink, Unprotected Transient Overpower and Unprotected Safe Shutdown Earthquake. (author)

Steady and transient regimes in hydropower plants

Science.gov (United States)

Gajic, A.

2013-12-01

Hydropower plant that has been in operation for about 30 years has to be reconstructed. They have already installed 12 Kaplan turbines, the largest in the world at that time. The existing CAM relationship was determined based on hydraulic model tests and checked by efficiency on-site tests. It was also tested based on turbine bearing vibrations. In order to discover vibrations and long cracks on stay vanes detailed on-site measurements were performed. Influence of the modification of the trailing edges on the dynamic stresses of the stay vanes is also shown. In order to improve power output transient regimes were analyzed, both experimentally and numerically. Reversible hydropower plant, a pioneer in Europe since it was the first Pump storage power plant constructed with the highest head pump-turbines in the world. Analyses of transient regimes discover some problems with S-shaped characteristics coupled with non-symmetrical penstock.

Steady and transient regimes in hydropower plants

International Nuclear Information System (INIS)

Gajic, A

2013-01-01

Hydropower plant that has been in operation for about 30 years has to be reconstructed. They have already installed 12 Kaplan turbines, the largest in the world at that time. The existing CAM relationship was determined based on hydraulic model tests and checked by efficiency on-site tests. It was also tested based on turbine bearing vibrations. In order to discover vibrations and long cracks on stay vanes detailed on-site measurements were performed. Influence of the modification of the trailing edges on the dynamic stresses of the stay vanes is also shown. In order to improve power output transient regimes were analyzed, both experimentally and numerically. Reversible hydropower plant, a pioneer in Europe since it was the first Pump storage power plant constructed with the highest head pump-turbines in the world. Analyses of transient regimes discover some problems with S-shaped characteristics coupled with non-symmetrical penstock

Verification and validation of COBRA-SFS transient analysis capability

International Nuclear Information System (INIS)

Rector, D.R.; Michener, T.E.; Cuta, J.M.

1998-05-01

This report provides documentation of the verification and validation testing of the transient capability in the COBRA-SFS code, and is organized into three main sections. The primary documentation of the code was published in September 1995, with the release of COBRA-SFS, Cycle 2. The validation and verification supporting the release and licensing of COBRA-SFS was based solely on steady-state applications, even though the appropriate transient terms have been included in the conservation equations from the first cycle. Section 2.0, COBRA-SFS Code Description, presents a capsule description of the code, and a summary of the conservation equations solved to obtain the flow and temperature fields within a cask or assembly model. This section repeats in abbreviated form the code description presented in the primary documentation (Michener et al. 1995), and is meant to serve as a quick reference, rather than independent documentation of all code features and capabilities. Section 3.0, Transient Capability Verification, presents a set of comparisons between code calculations and analytical solutions for selected heat transfer and fluid flow problems. Section 4.0, Transient Capability Validation, presents comparisons between code calculations and experimental data obtained in spent fuel storage cask tests. Based on the comparisons presented in Sections 2.0 and 3.0, conclusions and recommendations for application of COBRA-SFS to transient analysis are presented in Section 5.0

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

Response of air cleaning system dampers and blowers to simulated tornado transients

International Nuclear Information System (INIS)

Gregory, W.; Idar, E.; Smith, P.; Hensel, E.; Smith, E.

1985-01-01

The effects of tornado-like pressure transients upon dampers and blowers in nuclear air cleaning systems were studied. For the dampers pressure drop as a function of flow rate was obtained and an empirical relationship developed. Transient response was examined for several types of dampers, as was structural integrity. Both centrifugal and axi-vane blowers were tested and transient characteristic curves were generated in outrunning and backflow situations. The transient characteristic curves do not necessarily match the quasi-steady characteristic curves

Empirical method to calculate Clinch River Breeder Reactor (CRBR) inlet plenum transient temperatures

International Nuclear Information System (INIS)

Howarth, W.L.

1976-01-01

Sodium flow enters the CRBR inlet plenum via three loops or inlets. An empirical equation was developed to calculate transient temperatures in the CRBR inlet plenum from known loop flows and temperatures. The constants in the empirical equation were derived from 1/4 scale Inlet Plenum Model tests using water as the test fluid. The sodium temperature distribution was simulated by an electrolyte. Step electrolyte transients at 100 percent model flow were used to calculate the equation constants. Step electrolyte runs at 50 percent and 10 percent flow confirmed that the constants were independent of flow. Also, a transient was tested which varied simultaneously flow rate and electrolyte. Agreement of the test results with the empirical equation results was good which verifies the empirical equation

Newborn hearing screening with transient evoked otoacoustic emissions and automatic auditory brainstem response

Directory of Open Access Journals (Sweden)

Renata Mota Mamede de Carvallo

2008-09-01

Full Text Available Objective: The aim of the present investigation was to check Transient Evoked Otoacoustic Emissions and Automatic Auditory Brainstem Response tests applied together in regular nurseries and Newborn Intensive Care Units (NICU, as well as to describe and compare the results obtained in both groups. Methods: We tested 150 newborns from regular nurseries and 70 from NICU. Rresults: The newborn hearing screening results using Transient Evoked Otoacoustic Emissions and Automatic Auditory Brainstem Response tests could be applied to all babies. The “pass” result for the group of babies from the nursery was 94.7% using Transient Evoked Otoacoustic Emissions and 96% using Automatic Auditory Brainstem Response. The newborn intensive care unit group obtained 87.1% on Transient Evoked Otoacoustic Emissions and 80% on the Automatic Auditory Brainstem Response, and there was no statistical difference between the procedures when the groups were evaluated individually. However, comparing the groups, Transient Evoked Otoacoustic Emissions were presented in 94.7% of the nursery babies and in 87.1% in the group from the newborn intensive care unit. Considering the Automatic Auditory Brainstem Response, we found 96 and 87%, respectively. Cconclusions: Transient Evoked Otoacoustic Emissions and Automatic Auditory Brainstem Response had similar “pass” and “fail” results when the procedures were applied to neonates from the regular nursery, and the combined tests were more precise to detect hearing impairment in the newborn intensive care unit babies.

A pilot application of the RELAP file to the steady state and transient analysis of a test section inside the BR2 reactor

International Nuclear Information System (INIS)

Ferri, M. G.; D'Auria, F.; Forasassi, G.; Giot, M.

2000-01-01

BR2 is a material test reactor sited in the Belgian Nuclear Research Centre in Mol. The main research programs carried out in BR2 are related to the safety of nuclear reactor structural materials and fuels, in normal and accidental conditions, plant lifetime evaluation and ageing of components. In this framework, a computer program that allows the performance of detailed, steady state analysis of several kinds of in-pile sections with an axisymmetrical geometry has been developed. Furthermore, comparing its results with those of the well known, extensively used, Relap5/Mod 3.2 code on a test problem has validated this program. This was performed in three steps: 1. modalisation development of a subsystem of a typical in-pile section. 2. steady state analysis and comparison with the above-mentioned program. 3. transient simulation of the same system; the considered transient consists of a loss of coolant flow. (author)

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

Analytical prediction and experimental verification of reactor safety system injection transient

International Nuclear Information System (INIS)

Roy, B.N.; Nomm, E.

1991-01-01

This paper describes the computer code that was developed for thermal hydraulic transient analysis of mixed phase fluid system and the flow tests that were carried out to validate the Code. A full scale test facility was designed to duplicate the Supplementary Shutdown System (SSS) of Savannah River Production Reactors. Several steady state and dynamic flow tests were conducted simulating the actual reactor injection transients. A dynamic multiphase fluid flow code was developed and validated with experimental results and utilized for system performance predictions and development of technical specifications for reactors. 3 refs

Switching transients in a superconducting coil

International Nuclear Information System (INIS)

Owen, E.W.; Shimer, D.W.

1983-01-01

A study is made of the transients caused by the fast dump of large superconducting coils. Theoretical analysis, computer simulation, and actual measurements are used. Theoretical analysis can only be applied to the simplest of models. In the computer simulations two models are used, one in which the coil is divided into ten segments and another in which a single coil is employed. The circuit breaker that interrupts the current to the power supply, causing a fast dump, is represented by a time and current dependent conductance. Actual measurements are limited to measurements made incidental to performance tests on the MFTF Yin-yang coils. It is found that the breaker opening time is the critical factor in determining the size and shape of the transient. Instantaneous opening of the breaker causes a lightly damped transient with large amplitude voltages to ground. Increasing the opening time causes the transient to become a monopulse of decreasing amplitude. The voltages at the external terminals are determined by the parameters of the external circuit. For fast opening times the frequency depends on the dump resistor inductance, the circuit capacitance, and the amplitude on the coil current. For slower openings the dump resistor inductance and the current determine the amplitude of the voltage to ground at the terminals. Voltages to ground are less in the interior of the coil, where transients related to the parameters of the coil itself are observed

Fission gas behavior in mixed-oxide fuel during transient overpower

International Nuclear Information System (INIS)

Randklev, E.H.; Treibs, H.A.; Mastel, B.; Baldwin, D.L.

1979-01-01

Fission gas behavior can be important in determining fuel pin and core performance during a reactor transient. The results are presented of examinations characterizing the changes in microstructural distribution and retention of fission gas in fuel for a series of transient overpower (50 cents/s) tested mixed-oxide fuel pins and their steady state siblings

Relationship between Cerebral Microbleeds and Liver Stiffness Determined by Transient Elastography.

Directory of Open Access Journals (Sweden)

Young Dae Kim

Full Text Available Liver fibrosis is a multifactorial disease that can affect the development of cerebral small vessel diseases (SVDs including cerebral microbleeds (CMBs, leukoaraiosis, and silent infarctions. Transient elastography can accurately assess the degree of liver fibrosis by measuring liver stiffness (LS. In the present study, we investigated the association between SVDs and LS values.We recruited 300 participants (mean age 56 years, 170 men who underwent a comprehensive medical health check-up between January 2011 and December 2012. Transient elastography was taken on the right lobe of the liver through intercostal space with patients lying in the dorsal decubitus position with the right arm in maximal abduction. Mild and significant fibrosis were defined as LS values >5.6 and >8.0 kPa, respectively. The presence of each SVD was determined using the FLAIR, GRE MR imaging as well as T1-, T2-weighted MR images. We tested whether the presence and burden of each type of SVD were different by LS values.Of the different types of SVDs, only the presence (p = 0.001 and number of CMBs (p8.0 kPa was an independent predictor of CMBs (odds ratio 6.079, 95% confidence interval 1.489-24.819, p = 0.012. However, leukoaraiosis and silent infarctions were not associated with LS values (all p>0.05.The degree of liver fibrosis, as assessed using transient elastography, was independently associated with the presence and burden of CMBs in healthy, asymptomatic participants. Understanding the link between the brain and liver may advance future research on the pathomechanisms of CMBs.

Tests of gravity with future space-based experiments

Science.gov (United States)

Sakstein, Jeremy

2018-03-01

Future space-based tests of relativistic gravitation—laser ranging to Phobos, accelerometers in orbit, and optical networks surrounding Earth—will constrain the theory of gravity with unprecedented precision by testing the inverse-square law, the strong and weak equivalence principles, and the deflection and time delay of light by massive bodies. In this paper, we estimate the bounds that could be obtained on alternative gravity theories that use screening mechanisms to suppress deviations from general relativity in the Solar System: chameleon, symmetron, and Galileon models. We find that space-based tests of the parametrized post-Newtonian parameter γ will constrain chameleon and symmetron theories to new levels, and that tests of the inverse-square law using laser ranging to Phobos will provide the most stringent constraints on Galileon theories to date. We end by discussing the potential for constraining these theories using upcoming tests of the weak equivalence principle, and conclude that further theoretical modeling is required in order to fully utilize the data.

Testing IH Instrumentation: Analysis of 1996-1998 Tank Ventilation Data in Terms of Characterizing a Transient Release

International Nuclear Information System (INIS)

Droppo, James G.

2004-01-01

An analysis is conducted of the 1996-1998 Hanford tank ventilation studies of average ventilation rates to help define characteristics of shorter term releases. This effort is being conducted as part of the design of tests of Industrial Hygiene's (IH) instrumentation ability to detect transient airborne plumes from tanks using current deployment strategies for tank operations. This analysis has improved our understanding of the variability of hourly average tank ventilation processes. However, the analysis was unable to discern the relative importance of emissions due to continuous releases and short-duration bursts of material. The key findings are as follows: (1) The ventilation of relatively well-sealed, passively ventilated tanks appears to be driven by a combination of pressure, buoyancy, and wind influences. The results of a best-fit analysis conducted with a single data set provide information on the hourly emission variability that IH instrumentation will need to detect. (2) Tank ventilation rates and tank emission rates are not the same. The studies found that the measured infiltration rates for a single tank are often a complex function of air exchanges between tanks and air exchanges with outdoor air. This situation greatly limits the usefulness of the ventilation data in defining vapor emission rates. (3) There is no evidence in the data to discern if the routine tank vapor releases occur over a short time (i.e., a puff) or over an extended time (i.e., continuous releases). Based on this analysis of the tank ventilation studies, it is also noted that (1) the hourly averaged emission peaks from the relatively well-sealed passively-vented tanks (such as U-103) are not a simple function of one meteorological parameter--but the peaks often are the result of the coincidence of temporal maximums in pressure, temperature, and wind influences and (2) a mechanistic combination modeling approach and/or field studies may be necessary to understand the short

Thermal-hydraulic tests with out-of-pile test facility for BOCA development

International Nuclear Information System (INIS)

Kitagishi, Shigeru; Aoyama, Masashi; Tobita, Masahiro; Inaba, Yoshitomo; Yamaura, Takayuki

2012-01-01

The fuel transient test facility was prepared for power ramping tests of light-water-reactor (LWR) fuels in the Japan Materials Testing Reactor (JMTR) under a contract project with the Nuclear Industrial Safety Agent (NISA) of the Ministry of Economy, Trade and Industry (METI). It is necessary to develop high accuracy analysis procedure for power ramping tests after restart of the JMTR. The out-of-pile test facility to simulate thermal-hydraulic conditions of the fuel transient test facility was therefore developed. Applicability of the analysis code ACE-3D was examined for thermal-hydraulic analysis of power ramping tests for 10x10 BWR fuels by the fuel transient test facility. As the results, the calculated temperature was 304°C in comparison with measured value of 304.9-317.4°C in the condition of 600 W/cm. There is a bright prospect of high accuracy power ramping tests by the fuel transient test facility in JMTR. (author)

Development of DCC software dynamic test facility: past and future

International Nuclear Information System (INIS)

McDonald, A.M.; Thai, N.D.; Buijs, W.J.

1996-01-01

This paper describes a test facility for future dynamic testing of DCC software used in the control computers of CANDU nuclear power stations. It is a network of three computers: the DCC emulator, the dynamic CANDU plant simulator and the testing computer. Shared network files are used for input/output data exchange between computers. The DCC emulator runs directly on the binary image of the DCC software. The dynamic CANDU plant simulator accepts control signals from the DCC emulator and returns realistic plant behaviour. The testing computer accepts test scripts written in AECL Test Language. Both dynamic test and static tests may be performed on the DCC software to verify control program outputs and dynamic responses. (author)

RADIO TRANSIENTS FROM ACCRETION-INDUCED COLLAPSE OF WHITE DWARFS

International Nuclear Information System (INIS)

Moriya, Takashi J.

2016-01-01

We investigate observational properties of accretion-induced collapse (AIC) of white dwarfs (WDs) in radio frequencies. If AIC is triggered by accretion from a companion star, a dense circumstellar medium can be formed around the progenitor system. Then, the ejecta from AIC collide with the dense circumstellar medium, creating a strong shock. The strong shock can produce synchrotron emission that can be observed in radio frequencies. Even if AIC occurs as a result of WD mergers, we argue that AIC may cause fast radio bursts (FRBs) if a certain condition is satisfied. If AIC forms neutron stars (NSs) that are so massive that rotation is required to support themselves (i.e., supramassive NSs), the supramassive NSs may immediately lose their rotational energy by the r-mode instability and collapse to black holes. If the collapsing supramassive NSs are strongly magnetized, they may emit FRBs, as previously proposed. The AIC radio transients from single-degenerate systems may be detected in future radio transient surveys like the Very Large Array Sky Survey or the Square Kilometer Array transient survey. Because AIC has been proposed as a source of gravitational waves (GWs), GWs from AIC may be accompanied by radio-bright transients that can be used to confirm the AIC origin of observed GWs.

RADIO TRANSIENTS FROM ACCRETION-INDUCED COLLAPSE OF WHITE DWARFS

Energy Technology Data Exchange (ETDEWEB)

Moriya, Takashi J., E-mail: takashi.moriya@nao.ac.jp [Division of Theoretical Astronomy, National Astronomical Observatory of Japan, National Institutes of Natural Sciences, 2-21-1 Osawa, Mitaka, Tokyo 181-8588 (Japan)

2016-10-20

We investigate observational properties of accretion-induced collapse (AIC) of white dwarfs (WDs) in radio frequencies. If AIC is triggered by accretion from a companion star, a dense circumstellar medium can be formed around the progenitor system. Then, the ejecta from AIC collide with the dense circumstellar medium, creating a strong shock. The strong shock can produce synchrotron emission that can be observed in radio frequencies. Even if AIC occurs as a result of WD mergers, we argue that AIC may cause fast radio bursts (FRBs) if a certain condition is satisfied. If AIC forms neutron stars (NSs) that are so massive that rotation is required to support themselves (i.e., supramassive NSs), the supramassive NSs may immediately lose their rotational energy by the r-mode instability and collapse to black holes. If the collapsing supramassive NSs are strongly magnetized, they may emit FRBs, as previously proposed. The AIC radio transients from single-degenerate systems may be detected in future radio transient surveys like the Very Large Array Sky Survey or the Square Kilometer Array transient survey. Because AIC has been proposed as a source of gravitational waves (GWs), GWs from AIC may be accompanied by radio-bright transients that can be used to confirm the AIC origin of observed GWs.

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,

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

Deep knowledge expert system for diagnosis of multiple-failure severe transients in nuclear power plant

International Nuclear Information System (INIS)

Martin, R.P.; Nassersharif, B.

1987-01-01

TAMUS (Transient Analysis of MUltiple-failure Simulations) is a prototype expert system which is the result of a project investigating and implementing event confidence-levels (used by reactor safety experts in reactor transient analysis) in the form of an expert system. Currently, TAMUS is designed to diagnose reactor transients by analyzing simulated sensor and plant thermal hydraulic information from a system simulation. TAMUS uses a knowledge base of existing emergency nuclear plant operating guidelines and detailed thermal-hydraulic calculating results correlated to confidence-levels. TAMUS can diagnose a number of reactor transients (for example, loss-of-coolant accidents, steam-generator-tube ruptures, loss-of-offsite power, etc.). Future work includes the expansion of the knowledge base and improvement of the deep-knowledge qualitative models

Transient Seepage for Levee Engineering Analyses

Science.gov (United States)

Tracy, F. T.

2017-12-01

Historically, steady-state seepage analyses have been a key tool for designing levees by practicing engineers. However, with the advances in computer modeling, transient seepage analysis has become a potentially viable tool. A complication is that the levees usually have partially saturated flow, and this is significantly more complicated in transient flow. This poster illustrates four elements of our research in partially saturated flow relating to the use of transient seepage for levee design: (1) a comparison of results from SEEP2D, SEEP/W, and SLIDE for a generic levee cross section common to the southeastern United States; (2) the results of a sensitivity study of varying saturated hydraulic conductivity, the volumetric water content function (as represented by van Genuchten), and volumetric compressibility; (3) a comparison of when soils do and do not exhibit hysteresis, and (4) a description of proper and improper use of transient seepage in levee design. The variables considered for the sensitivity and hysteresis studies are pore pressure beneath the confining layer at the toe, the flow rate through the levee system, and a levee saturation coefficient varying between 0 and 1. Getting results for SEEP2D, SEEP/W, and SLIDE to match proved more difficult than expected. After some effort, the results matched reasonably well. Differences in results were caused by various factors, including bugs, different finite element meshes, different numerical formulations of the system of nonlinear equations to be solved, and differences in convergence criteria. Varying volumetric compressibility affected the above test variables the most. The levee saturation coefficient was most affected by the use of hysteresis. The improper use of pore pressures from a transient finite element seepage solution imported into a slope stability computation was found to be the most grievous mistake in using transient seepage in the design of levees.

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 simulation for a real-time operator advisor expert system

International Nuclear Information System (INIS)

Jakubowski, T.; Hajek, B.K.; Miller, D.W.; Bhatnagar, R.

1990-01-01

An Operator Advisor (OA) consisting of four integrated expert systems has been under development at The Ohio State University since 1985. The OA, designed for a General Electric BWR-6 plant, has used the Perry Nuclear Power Plants full scope simulator near Cleveland, Ohio (USA) as the reference plant. The primary goal of this development has been to provide a single system which not only performs monitoring and diagnosis functions, but also provides fault mitigation procedures to the operator, monitors the performance of these procedures and provides backup procudures should the initial ones fail. To test the system off line, a transient event simulation methodology has been developed. The simulator employs event scenarios from the Perry simulator. Scenarios are selected to test both inter- and intra-modular system behavior and response and to verify the consistency and accuracy of the knowledge base. This paper describes the OA architecture and design and the transient simulation. A discussion of a sample scenario test, including the rationale for scenario selection, is included as an example. The results of the testing demonstrate the value of off line transient simulations in that they verify system operation and help to identify characteristics requiring further improvements

Bright PanSTARRS Nuclear Transients – what are they?

Directory of Open Access Journals (Sweden)

Smartt S.

2012-12-01

Full Text Available We present an initial analysis of 49 bright transients occurring in the nuclei of galaxies with no previous known Active Galactic Nucleus (AGN. They have been discovered as part of the PanSTARRs 3π survey, and followed up with the Liverpool Telescope. Based on colours, light curve shape, and a small number with optical spectra, these transients seem to fall into three groups. Red/fast transients are nuclear supernovae of various types. Some bright nuclear transients are blue and decay on a timescale of a few months; these may be candidates for tidal disruption events. However most of the events we have found are blue and are either still rising or decaying slowly, on a timescale of years; the few spectra we have show AGN at z ∼ 1. We argue that these transients are background AGN microlensed by stars in foreground galaxies by a factor 10–100. Monitoring such events gives us very promising prospects for measuring the structure of AGN and so testing current theories.

Peach Bottom Turbine Trip Simulations with RETRAN Using INER/TPC BWR Transient Analysis Method

International Nuclear Information System (INIS)

Kao Lainsu; Chiang, Show-Chyuan

2005-01-01

The work described in this paper is benchmark calculations of pressurization transient turbine trip tests performed at the Peach Bottom boiling water reactor (BWR). It is part of an overall effort in providing qualification basis for the INER/TPC BWR transient analysis method developed for the Kuosheng and Chinshan plants. The method primarily utilizes an advanced system thermal hydraulics code, RETRAN02/MOD5, for transient safety analyses. Since pressurization transients would result in a strong coupling effect between core neutronic and system thermal hydraulics responses, the INER/TPC method employs the one-dimensional kinetic model in RETRAN with a cross-section data library generated by the Studsvik-CMS code package for the transient calculations. The Peach Bottom Turbine Trip (PBTT) tests, including TT1, TT2, and TT3, have been successfully performed in the plant and assigned as standards commonly for licensing method qualifications for years. It is an essential requirement for licensing purposes to verify integral capabilities and accuracies of the codes and models of the INER/TPC method in simulating such pressurization transients. Specific Peach Bottom plant models, including both neutronics and thermal hydraulics, are developed using modeling approaches and experiences generally adopted in the INER/TPC method. Important model assumptions in RETRAN for the PBTT test simulations are described in this paper. Simulation calculations are performed with best-estimated initial and boundary conditions obtained from plant test measurements. The calculation results presented in this paper demonstrate that the INER/TPC method is capable of calculating accurately the core and system transient behaviors of the tests. Excellent agreement, both in trends and magnitudes between the RETRAN calculation results and the PBTT measurements, shows reliable qualifications of the codes/users/models involved in the method. The RETRAN calculated peak neutron fluxes of the PBTT

Analysis, by Relap5 code, of boron dilution phenomena in a Small Break Loca Transient, performed in PKL III E 2.2 test

International Nuclear Information System (INIS)

Rizzo, G.; Vella, G.

2007-01-01

The present work is finalized to investigate the E2.2 thermal-hydraulics transient of the PKL III facility, which is a scaled reproduction of a typical German PWR, operated by FRAMATOME-ANP in Erlangen, Germany, within the framework of an international cooperation (OECD/SETH project). The main purpose of the project is to study boron dilution events in Pressurized Water Reactors and to contribute to the assessment of thermal-hydraulic system codes like Relap5. The experimental test PKL III E2.2 investigates the behavior of a typical PWR after a Small Break Loss Of Coolant Accident (SB-LOCA) in a cold leg and an immediate injection of borated water in two cold legs. The main purpose of this work is to simulate the PKL III test facility and particularly its experimental transient by Relap5 system code. The adopted nodalization, already available at Department of Nuclear Engineering (DIN), has been reviewed and applied with an accurate analysis of the experimental test parameters. The main result relies in a good agreement of calculated data with experimental measures for a number of main important variables. (author)

Transient electromagnetic analysis in tokamaks using TYPHOON code

International Nuclear Information System (INIS)

Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

1996-01-01

The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

PC operated acoustic transient spectroscopy of deep levels in MIS structures

International Nuclear Information System (INIS)

Bury, P.; Jamnicky, I.

1996-01-01

A new version of acoustic deep-level transient spectroscopy is presented to study the traps at the insulator-semiconductor interface. The acoustic deep-level transient spectroscopy uses an acoustoelectric response signal produced by the MIS structure interface when a longitudinal acoustic wave propagates through a structure. The acoustoelectric response signal is extremely sensitive to external conditions of the structure and reflects any changes in the charge distribution, connected also with charged traps. In comparison with previous version of acoustic deep-level transient spectroscopy that closely coincides with the principle of the original deep-level transient spectroscopy technique, the present technique is based on the computer-evaluated isothermal transients and represents an improved, more efficient and time saving technique. Many tests on the software used for calculation as well as on experimental setup have been performed. The improved acoustic deep-level transient spectroscopy method has been applied for the Si(p) MIS structures. The deep-level parameters as activation energy and capture cross-section have been determined. (authors)

Investigating the impact of in-vehicle transients on diesel soot emissions

Directory of Open Access Journals (Sweden)

Filipi Zoran

2008-01-01

Full Text Available This paper describes development of a test cell setup for concurrent running of a real engine and a simulation of the vehicle system, and its use for investigating highly-dynamic engine-in-vehicle operation and its effect on diesel engine emissions. Running an engine in the test cell under conditions experienced in the vehicle enables acquiring detailed insight into dynamic interactions between power train sub-systems, and the impact of it on fuel consumption and transient emissions. This type of data may otherwise be difficult and extremely costly to obtain from a vehicle prototype test. In particular, engine system response during critical transients and the effect of transient excursions on emissions are investigated using advanced, fast-response test instrumentation and emissions analyzers. Main enablers of the work include the highly dynamic AC electric dynamometer with the accompanying computerized control system and the computationally efficient simulation of the driveline/vehicle system. The latter is developed through systematic energy-based proper modeling that tailors the virtual model to capture critical powertrain transients while running in real time. Coupling the real engine with the virtual driveline/vehicle offers a chance to easily modify vehicle parameters, and even study different power train configurations. In particular, the paper describes the engine-in-the-loop study of a V-8, 6l engine coupled to a virtual 4´4 off road vehicle. This engine is considered as a high-performance option for this truck and the real prototype of the complete vehicle does not exist yet. The results shed light on critical transients in a conventional powertrain and their effect on NOx and soot emissions. Measurements demonstrate very large spikes of particulate concentration at the initiation of vehicle acceleration events. Characterization of transients and their effect on particulate emission provides a basis for devising engine-level or

Transient characteristics of parallel running of the 20kVA superconducting synchronous generator and a conventional one

International Nuclear Information System (INIS)

Nitta, T.; Okada, T.

1989-01-01

This paper describes electrical transient characteristics of parallel running of the 20kVA superconducting synchronous generator and a conventional one. In the experimental power system, the superconducting generator is connected through reactors (artificial transmission lines) to a regional power system (infinite bus) and the conventional generator (20kVA) is connected to the terminal of the superconducting generator. Several tests were performed in order to consider the transient behavior of superconducting generator (SCG) in the power system. The items of the tests are synchronous closing test, loss of synchronism test and disconnecting and reclosing test. From the experimental results, it can be said that by installing SCG in power systems, voltage stability and power system stability can be improved in transient states as well as in steady states and the variation of armature current of SCG during a transient period is extremely larger than that of the conventional one. The transient analysis by a computer simulation was also carried out for the experiments. The simulation results are in good agreement with the experimental ones

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.

Validation and comparison of two-phase flow modeling capabilities of CFD, sub channel and system codes by means of post-test calculations of BFBT transient tests

Energy Technology Data Exchange (ETDEWEB)

Jaeger, Wadim; Manes, Jorge Perez; Imke, Uwe; Escalante, Javier Jimenez; Espinoza, Victor Sanchez, E-mail: victor.sanchez@kit.edu

2013-10-15

Highlights: • Simulation of BFBT turbine and pump transients at multiple scales. • CFD, sub-channel and system codes are used for the comparative study. • Heat transfer models are compared to identify difference between the code predictions. • All three scales predict results in good agreement to experiment. • Sub cooled boiling models are identified as field for future research. -- Abstract: The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in the validation and qualification of modern thermo hydraulic simulations tools at various scales. In the present paper, the prediction capabilities of four codes from three different scales – NEPTUNE{sub C}FD as fine mesh computational fluid dynamics code, SUBCHANFLOW and COBRA-TF as sub channels codes and TRACE as system code – are assessed with respect to their two-phase flow modeling capabilities. The subject of the investigations is the well-known and widely used data base provided within the NUPEC BFBT benchmark related to BWRs. Void fraction measurements simulating a turbine and a re-circulation pump trip are provided at several axial levels of the bundle. The prediction capabilities of the codes for transient conditions with various combinations of boundary conditions are validated by comparing the code predictions with the experimental data. In addition, the physical models of the different codes are described and compared to each other in order to explain the different results and to identify areas for further improvements.

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

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

Transient chaos in a globally coupled system of nearly conservative Hamiltonian Duffing oscillators

International Nuclear Information System (INIS)

Sabarathinam, S.; Thamilmaran, K.

2015-01-01

Highlights: •We have examined transient chaos in globally coupled oscillators. •We analyze transient chaos using new techniques. •We give experimental confirmation of transient chaos. -- Abstract: In this work, transient chaos in a ring and globally coupled system of nearly conservative Hamiltonian Duffing oscillators is reported. The networks are formed by coupling of three, four and six Duffing oscillators. The nearly conservative Hamiltonian nature of the coupled system is proved by stability analysis. The transient phenomenon is confirmed through various numerical investigations such as recurrence analysis, 0–1 test and Finite Time Lyapunov Exponents. Further, the effect of damping and the average transient lifetime of three, four and six coupled schemes for randomly generated initial conditions have been analyzed. The experimental confirmation of transient chaos in an illustrative system of three ringly coupled Duffing oscillators is also presented

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-04-01

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

Application of statistical method for FBR plant transient computation

International Nuclear Information System (INIS)

Kikuchi, Norihiro; Mochizuki, Hiroyasu

2014-01-01

Highlights: • A statistical method with a large trial number up to 10,000 is applied to the plant system analysis. • A turbine trip test conducted at the “Monju” reactor is selected as a plant transient. • A reduction method of trial numbers is discussed. • The result with reduced trial number can express the base regions of the computed distribution. -- Abstract: It is obvious that design tolerances, errors included in operation, and statistical errors in empirical correlations effect on the transient behavior. The purpose of the present study is to apply above mentioned statistical errors to a plant system computation in order to evaluate the statistical distribution contained in the transient evolution. A selected computation case is the turbine trip test conducted at 40% electric power of the prototype fast reactor “Monju”. All of the heat transport systems of “Monju” are modeled with the NETFLOW++ system code which has been validated using the plant transient tests of the experimental fast reactor Joyo, and “Monju”. The effects of parameters on upper plenum temperature are confirmed by sensitivity analyses, and dominant parameters are chosen. The statistical errors are applied to each computation deck by using a pseudorandom number and the Monte-Carlo method. The dSFMT (Double precision SIMD-oriented Fast Mersenne Twister) that is developed version of Mersenne Twister (MT), is adopted as the pseudorandom number generator. In the present study, uniform random numbers are generated by dSFMT, and these random numbers are transformed to the normal distribution by the Box–Muller method. Ten thousands of different computations are performed at once. In every computation case, the steady calculation is performed for 12,000 s, and transient calculation is performed for 4000 s. In the purpose of the present statistical computation, it is important that the base regions of distribution functions should be calculated precisely. A large number of

Transient analysis of a variable speed rotary compressor

International Nuclear Information System (INIS)

Park, Youn Cheol

2010-01-01

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

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

Effects of transients in LIGO suspensions on searches for gravitational waves.

Science.gov (United States)

Walker, M; Abbott, T D; Aston, S M; González, G; Macleod, D M; McIver, J; Abbott, B P; Abbott, R; Adams, C; Adhikari, R X; Anderson, S B; Ananyeva, A; Appert, S; Arai, K; Ballmer, S W; Barker, D; Barr, B; Barsotti, L; Bartlett, J; Bartos, I; Batch, J C; Bell, A S; Betzwieser, J; Billingsley, G; Birch, J; Biscans, S; Biwer, C; Blair, C D; Bork, R; Brooks, A F; Ciani, G; Clara, F; Countryman, S T; Cowart, M J; Coyne, D C; Cumming, A; Cunningham, L; Danzmann, K; Da Silva Costa, C F; Daw, E J; DeBra, D; DeRosa, R T; DeSalvo, R; Dooley, K L; Doravari, S; Driggers, J C; Dwyer, S E; Effler, A; Etzel, T; Evans, M; Evans, T M; Factourovich, M; Fair, H; Fernández Galiana, A; Fisher, R P; Fritschel, P; Frolov, V V; Fulda, P; Fyffe, M; Giaime, J A; Giardina, K D; Goetz, E; Goetz, R; Gras, S; Gray, C; Grote, H; Gushwa, K E; Gustafson, E K; Gustafson, R; Hall, E D; Hammond, G; Hanks, J; Hanson, J; Hardwick, T; Harry, G M; Heintze, M C; Heptonstall, A W; Hough, J; Izumi, K; Jones, R; Kandhasamy, S; Karki, S; Kasprzack, M; Kaufer, S; Kawabe, K; Kijbunchoo, N; King, E J; King, P J; Kissel, J S; Korth, W Z; Kuehn, G; Landry, M; Lantz, B; Lockerbie, N A; Lormand, M; Lundgren, A P; MacInnis, M; Márka, S; Márka, Z; Markosyan, A S; Maros, E; Martin, I W; Martynov, D V; Mason, K; Massinger, T J; Matichard, F; Mavalvala, N; McCarthy, R; McClelland, D E; McCormick, S; McIntyre, G; Mendell, G; Merilh, E L; Meyers, P M; Miller, J; Mittleman, R; Moreno, G; Mueller, G; Mullavey, A; Munch, J; Nuttall, L K; Oberling, J; Oliver, M; Oppermann, P; Oram, Richard J; O'Reilly, B; Ottaway, D J; Overmier, H; Palamos, J R; Paris, H R; Parker, W; Pele, A; Penn, S; Phelps, M; Pierro, V; Pinto, I; Principe, M; Prokhorov, L G; Puncken, O; Quetschke, V; Quintero, E A; Raab, F J; Radkins, H; Raffai, P; Reid, S; Reitze, D H; Robertson, N A; Rollins, J G; Roma, V J; Romie, J H; Rowan, S; Ryan, K; Sadecki, T; Sanchez, E J; Sandberg, V; Savage, R L; Schofield, R M S; Sellers, D; Shaddock, D A; Shaffer, T J; Shapiro, B; Shawhan, P; Shoemaker, D H; Sigg, D; Slagmolen, B J J; Smith, B; Smith, J R; Sorazu, B; Staley, A; Strain, K A; Tanner, D B; Taylor, R; Thomas, M; Thomas, P; Thorne, K A; Thrane, E; Torrie, C I; Traylor, G; Tuyenbayev, D; Vajente, G; Valdes, G; van Veggel, A A; Vecchio, A; Veitch, P J; Venkateswara, K; Vo, T; Vorvick, C; Ward, R L; Warner, J; Weaver, B; Weiss, R; Weßels, P; Willke, B; Wipf, C C; Worden, J; Wu, G; Yamamoto, H; Yancey, C C; Yu, Hang; Yu, Haocun; Zhang, L; Zucker, M E; Zweizig, J

2017-12-01

This paper presents an analysis of the transient behavior of the Advanced LIGO (Laser Interferometer Gravitational-wave Observatory) suspensions used to seismically isolate the optics. We have characterized the transients in the longitudinal motion of the quadruple suspensions during Advanced LIGO's first observing run. Propagation of transients between stages is consistent with modeled transfer functions, such that transient motion originating at the top of the suspension chain is significantly reduced in amplitude at the test mass. We find that there are transients seen by the longitudinal motion monitors of quadruple suspensions, but they are not significantly correlated with transient motion above the noise floor in the gravitational wave strain data, and therefore do not present a dominant source of background noise in the searches for transient gravitational wave signals. Using the suspension transfer functions, we compared the transients in a week of gravitational wave strain data with transients from a quadruple suspension. Of the strain transients between 10 and 60 Hz, 84% are loud enough that they would have appeared above the sensor noise in the top stage quadruple suspension monitors if they had originated at that stage at the same frequencies. We find no significant temporal correlation with the suspension transients in that stage, so we can rule out suspension motion originating at the top stage as the cause of those transients. However, only 3.2% of the gravitational wave strain transients are loud enough that they would have been seen by the second stage suspension sensors, and none of them are above the sensor noise levels of the penultimate stage. Therefore, we cannot eliminate the possibility of transient noise in the detectors originating in the intermediate stages of the suspension below the sensing noise.

Transient leak detection in crude oil pipelines

Energy Technology Data Exchange (ETDEWEB)

Beushausen, R.; Tornow, S.; Borchers, H. [Nord-West Oelleitung, Wilhelmshaven (Germany); Murphy, K.; Zhang, J. [Atmos International Ltd., Manchester (United Kingdom)

2004-07-01

Nord-West Oelleitung (NWO) operates 2 crude oil pipelines from Wilhemshaven to Koln and Hamburg respectively. German regulations for transporting flammable substances stipulate that 2 independent continuously working procedures be used to detect leaks. Leak detection pigs are used routinely to complement the surveillance system. This paper described the specific issues of transient leak detection in crude oil pipelines. It was noted that traditional methods have failed to detect leaks that occur immediately after pumps are turned on or off because the pressure wave generated by the transient dominates the pressure wave that results from the leak. Frequent operational changes in a pipeline are often accompanied by an increased number of false alarms and failure to detect leaks due to unsteady operations. NWO therefore decided to have the Atmos statistical pipeline leak detection (SPLD) system installed on their pipelines. The key to the SPLD system is the sequential probability ratio test. Comprehensive data validation is performed following reception of pipeline data from the supervisory control and data acquisition (SCADA) system. The validated data is then used to calculate the corrected flow imbalance, which is fed into the SPRT to determine if there is an increase in the flow imbalance. Pattern recognition is then used to distinguish a leak from operational changes. The SPLD is unique because it uses 3 computational pipeline monitoring methods simultaneously, namely modified volume balance, statistical analysis, and pressure and flow monitoring. The successful installation and testing of the SPLD in 2 crude oil pipelines was described along with the main difficulties associated with transient leaks. Field results were presented for both steady-state and transient conditions. 5 refs., 2 tabs., 16 figs.

Transient hyperthyroidism of hyperemesis gravidarum.

Science.gov (United States)

Tan, Jackie Y L; Loh, Keh Chuan; Yeo, George S H; Chee, Yam Cheng

2002-06-01

To characterise the clinical, biochemical and thyroid antibody profile in women with transient hyperthyroidism of hyperemesis gravidarum. Prospective observational study. Hospital inpatient gynaecological ward. Women admitted with hyperemesis gravidarum and found to have hyperthyroidism. Fifty-three women were admitted with hyperemesis gravidarum and were found to have hyperthyroidism. Each woman was examined for clinical signs of thyroid disease and underwent investigations including urea, creatinine, electrolytes, liver function test, thyroid antibody profile and serial thyroid function test until normalisation. Gestation at which thyroid function normalised, clinical and thyroid antibody profile and pregnancy outcome (birthweight, gestation at delivery and Apgar score at 5 minutes). Full data were available for 44 women. Free T4 levels normalised by 15 weeks of gestation in the 39 women with transient hyperthyroidism while TSH remained suppressed until 19 weeks of gestation. None of these women were clinically hyperthyroid. Thyroid antibodies were not found in most of them. Median birthweight in the infants of mothers who experienced weight loss of > 5% of their pre-pregnancy weight was lower compared with those of women who did not (P = 0.093). Five women were diagnosed with Graves' disease based on clinical features and thyroid antibody profile. In transient hyperthyroidism of hyperemesis gravidarum, thyroid function normalises by the middle of the second trimester without anti-thyroid treatment. Clinically overt hyperthyroidism and thyroid antibodies are usually absent. Apart from a non-significant trend towards lower birthweights in the infants of mothers who experienced significant weight loss, pregnancy outcome was generally good. Routine assessment of thyroid function is unnecessary for women with hyperemesis gravidarum in the absence of any clinical features of hyperthyroidism.

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.

SPIRITS: SPitzer InfraRed Intensive Transients Survey

Science.gov (United States)

Kasliwal, Mansi; Jencson, Jacob; Lau, Ryan; Masci, Frank; Helou, George; Williams, Robert; Bally, John; Bond, Howard; Whitelock, Patricia; Cody, Ann Marie; Gehrz, Robert; Tinyanont, Samaporn; Smith, Nathan; Surace, Jason; Armus, Lee; Cantiello, Matteo; Langer, Norbert; Levesque, Emily; Mohamed, Shazrene; Ofek, Eran; Parthasarathy, Mudumba; van Dyk, Schuyler; Boyer, Martha; Phillips, Mark; Hsiao, Eric; Morrell, Nidia; Perley, Dan; Gonzalez, Consuelo; Contreras, Carlos; Jones, Olivia; Ressler, Michael; Adams, Scott; Moore, Anna; Cook, David; Fox, Ori; Johansson, Joel; Khan, Rubab; Monson, Andrew; Hankins, Matthew; Goldman, Steven; Jacob, Jencson

2018-05-01

Spitzer is pioneering a systematic exploration of the dynamic infrared sky. Our SPitzer InfraRed Intensive Transients Survey (SPIRITS) has already discovered 78 explosive transients and 2457 eruptive variables. Of these 78 infrared transients, 60 are so red that they are devoid of optical counterparts and we call them SPRITEs (eSPecially Red Intermediate-luminosity Transient Events). The nature of SPRITEs is unknown and progress on deciphering the explosion physics depends on mid-IR spectroscopy. Multiple physical origins have been proposed including stellar merger, birth of a massive binary, electron capture supernova and stellar black hole formation. Hence, we propose a modest continuation of SPIRITS, focusing on discovering and monitoring SPRITEs, in preparation for follow-up with the James Webb Space Telescope (JWST). As the SPRITEs evolve and cool, the bulk of the emission shifts to longer wavelengths. MIRI aboard JWST will be the only available platform in the near future capable of characterizing SPRITEs out to 28 um. Specifically, the low resolution spectrometer would determine dust mass, grain chemistry, ice abundance and energetics to disentangle the proposed origins. The re-focused SPIRITS program consists of continued Spitzer monitoring of those 106 luminous galaxies that are known SPRITE hosts or are most likely to host new SPRITEs. Scaling from the SPIRITS discovery rate, we estimate finding 10 new SPRITEs and 2-3 new supernovae in Cycle 14. The SPIRITS team remains committed to extensive ground-based follow-up. The Spitzer observations proposed here are essential for determining the final fates of active SPRITEs as well as bridging the time lag between the current SPIRITS survey and JWST launch.

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

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

International Nuclear Information System (INIS)

Hall, D.G.; Watkins, J.C.

1987-01-01

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

1991-02-01

The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

Newborn hearing screening with transient evoked otoacoustic emissions and automatic auditory brainstem response

OpenAIRE

Renata Mota Mamede de Carvallo; Carla Gentile Matas; Isabela de Souza Jardim

2008-01-01

Objective: The aim of the present investigation was to check Transient Evoked Otoacoustic Emissions and Automatic Auditory Brainstem Response tests applied together in regular nurseries and Newborn Intensive Care Units (NICU), as well as to describe and compare the results obtained in both groups. Methods: We tested 150 newborns from regular nurseries and 70 from NICU. Rresults: The newborn hearing screening results using Transient Evoked Otoacoustic Emissions and Automatic Auditory Brainstem...

Transient global amnesia: neuropsychological dysfunction during attack and recovery in two "pure" cases.

OpenAIRE

Regard, M; Landis, T

1984-01-01

Two patients with transient global amnesia are reported. Comprehensive neuropsychological evaluation, during the amnesic episode, as well as follow-up examinations on memory were performed. The course of the amnesia was exemplified by two comparable memory tests in different modalities. Partial retrograde amnesia and complete anterograde amnesia were demonstrated during the transient episode. Objective recovery was found to be slower than subjectively experienced, but test performance was com...

Comparison of pressure transient response in intensely and sparsely fractured reservoirs

Energy Technology Data Exchange (ETDEWEB)

Johns, R.T.

1989-04-01

A comprehensive analytical model is presented to study the pressure transient behavior of a naturally fractured reservoir with a continuous matrix block size distribution. Geologically realistic probability density functions of matrix block size are used to represent reservoirs of varying fracture intensity and uniformity. Transient interporosity flow is assumed and interporosity skin is incorporated. Drawdown and interference pressure transient tests are investigated. The results show distinctions in the pressure response from intensely and sparsely fractured reservoirs in the absence of interporosity skin. Also, uniformly and nonuniformly fractured reservoirs exhibit distinct responses, irrespective of the degree of fracture intensity. The pressure response in a nonuniformly fractured reservoir with large block size variability, approaches a nonfractured (homogeneous) reservoir response. Type curves are developed to estimate matrix block size variability and the degree of fracture intensity from drawdown and interference well tests.

Transient stability index for online stability assessment and contingency evaluation

Energy Technology Data Exchange (ETDEWEB)

Ribbens-Pavella, M; Murthy, P G; Horward, J L; Carpentier, J L

1982-04-01

An on-line methodology is proposed for assessing the robustness of a power system from the point of view of transient stability, and a scalar expression, the transient stability index, is accordingly derived. The reliability and sensitivity of this index are tested by means of simulations for a number of power system cases. The index is shown to be appropriate for online stability assessment, contingency evaluation and preventive control. 14 refs.

Evaluation of Start Transient Oscillations with the J-2X Engine Gas Generator Assembly

Science.gov (United States)

Hulka, J. R.; Morgan, C. J.; Casiano, M. J.

2015-01-01

During development of the gas generator for the liquid oxygen/liquid hydrogen propellant J-2X rocket engine, distinctive and oftentimes high-amplitude pressure oscillations and hardware vibrations occurred during the start transient of nearly every workhorse gas generator assembly test, as well as during many tests of engine system hardware. These oscillations appeared whether the steady-state conditions exhibited stable behavior or not. They occurred similarly with three different injector types, and with every combustion chamber configuration tested, including chamber lengths ranging over a 5:1 range, several different nozzle types, and with or without a side branch line simulating a turbine spin start gas supply line. Generally, two sets of oscillations occurred, one earlier in the start transient and at higher frequencies, and the other almost immediately following and at lower frequencies. Multiple dynamic pressure measurements in the workhorse combustion chambers indicated that the oscillations were associated with longitudinal acoustic modes of the combustion chambers, with the earlier and higher frequency oscillation usually related to the second longitudinal acoustic mode and the later and lower frequency oscillation usually related to the first longitudinal acoustic mode. Given that several early development gas generator assemblies exhibited unstable behavior at frequencies near the first longitudinal acoustic modes of longer combustion chambers, the start transient oscillations are presumed to provide additional insight into the nature of the combustion instability mechanisms. Aspects of the steadystate oscillations and combustion instabilities from development and engine system test programs have been reported extensively in the three previous JANNAF Liquid Propulsion Subcommittee meetings (see references below). This paper describes the hardware configurations, start transient sequence operations, and transient and dynamic test data during the start

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

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

RELAP5/MOD2: for PWR transient analysis

International Nuclear Information System (INIS)

Ransom, V.H.

1983-01-01

RELAP5 is a light water reactor system transient simulation code for use in nuclear plant safety analysis. Development of a new version, RELAP5/MOD2, has been completed and will be released to the United States Nuclear Regulatory Commission during September of 1983. The new and improved modeling capability of RELAP5/MOD2 is described and some developmental assessment results are presented. The future plans for extension to severe accident modeling are briefly discussed

Trace coupled with PARCS benchmark against Leibstadt plant data during the turbine trip test

Energy Technology Data Exchange (ETDEWEB)

Sekhri, Abdelkrim; Baumann, Peter, E-mail: abdelkrim.sekhri@kkl.ch, E-mail: peter.Baumann@kkl.ch [KernkraftwerkLeibstadt AG, Leibstadt (Switzerland); Hidalga, Patricio; Morera, Daniel; Miro, Rafael; Barrachina, Teresa; Verdu, Gumersindo, E-mail: pathigar@etsii.upv.es, E-mail: dmorera@isirym.upv.es, E-mail: rmiro@isirym.upv.es, E-mail: tbarrachina@isirym.upv.es, E-mail: gverdu@isirym.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV), Valencia, (Spain). Institute for Industrial, Radiophysical and Environmental Safety

2013-07-01

In order to enhance the modeling of Nuclear Power Plant Leibstadt (KKL), the coupling of 3D neutron kinetics PARCS code with TRACE has been developed. To test its performance a complex transient of Turbine Trip has been simulated comparing the results with the existing plant data of Turbine Trip test. For this transient also Cross Sections have been generated and used by PARCS. The thermal-hydraulic TRACE model is retrieved from the already existing model. For the benchmarking the Turbine Trip transient has been simulated according to the test resulting in the closure of the turbine control valve (TCV) and the following opening of the bypass valve (TBV). This transient caused a pressure shock wave towards the Reactor Pressure Vessel (RPV) which provoked the decreasing of the void level and the consequent slight power excursion. The power control capacity of the system showed a good response with the procedure of a Selected Rod Insertion (SRI) and the recirculation loops performance which resulted in the proper thermal power reduction comparable to APRM data recorder from the plant. The comparison with plant data shows good agreement in general and assesses the performance of the coupled model. Due to this, it can be concluded that the coupling of PARCS and TRACE codes in addition with the Cross Section used works successfully for simulating the behavior of the reactor core during complex plant transients. Nevertheless the TRACE model shall be improved and the core neutronics corresponding to the test shall be used in the future to allow quantitative comparison between TRACE and plant recorded data. (author)

Fundamental study on thermo-hydraulic behaviors during power transient, 2

International Nuclear Information System (INIS)

Shinano, M.; Inoue, A.

1988-01-01

Thermo-hydraulic behaviors during power transient of nuclear reactors are studied. Boiling around test rod heated transiently forces to flow out liquid in the test section and generates high pressure pulse. In this study, it is investigated experimentally and analytically that magnitude of pressure pulse and energy conversion efficiency to the mechanical works in cases of fragmentation and non-fragmentation. In analysis, effects of increasing of heat transfer and of interaction area due to fragmentation is considered. Consequently, 1) magnitude of pressure pulse on fragmentation is about 10 times greater than that on non-fragmentation. 2) analytical model can show characteristics of fragmentation processes qualitatively. (author)

Finite strain transient creep of D16T alloy: identification and validation employing heterogeneous tests

Science.gov (United States)

Shutov, A. V.; Larichkin, A. Yu

2017-10-01

A cyclic creep damage model, previously proposed by the authors, is modified for a better description of the transient creep of D16T alloy observed in the finite strain range under rapidly changing stresses. The new model encompasses the concept of kinematic hardening, which allows us to account for the creep-induced anisotropy. The model kinematics is based on the nested multiplicative split of the deformation gradient, proposed by Lion. The damage evolution is accounted for by the classical Kachanov-Rabotnov approach. The material parameters are identified using experimental data on cyclic torsion of thick-walled samples with different holding times between load reversals. For the validation of the proposed material model, an additional experiment is analyzed. Although this additional test is not involved in the identification procedure, the proposed cyclic creep damage model describes it accurately.

Review and comparison of transient creep laws used for natural rock salt

International Nuclear Information System (INIS)

Herrmann, W.; Lauson, H.S.

1981-04-01

A number of creep laws are reviewed, which have been proposed to describe the transient creep of rock salt for use in design calculations of nuclear waste isolation and strategic petroleum reserve repositories. It is shown that they all have the same general form, and their ability to fit creep data for rock salt is tested. Four creep laws are found to fit the data for individual creep tests about equally well. Three of these include steady-state as well as transient creep, while the fourth, equivalent to power-law time hardening in the case of a creep test, does not. Extrapolations at constant stress and temperature of the three creep laws with steady-state creep essentially coincide for times longer than a few months, since the transient creep becomes negligible for such times. Power-law hardening, on the other hand, since it depends on time through a power less than one, predicts much smaller creep strains at very long times

Comprehensive embryo testing. Experts' opinions regarding future directions: an expert panel study on comprehensive embryo testing.

Science.gov (United States)

Hens, Kristien; Dondorp, Wybo J; Geraedts, Joep P M; de Wert, Guido M

2013-05-01

What do scientists in the field of preimplantation genetic diagnosis (PGD) and preimplantation genetic screening (PGS) consider to be the future direction of comprehensive embryo testing? Although there are many biological and technical limitations, as well as uncertainties regarding the meaning of genetic variation, comprehensive embryo testing will impact the IVF/PGD practice and a timely ethical reflection is needed. Comprehensive testing using microarrays is currently being introduced in the context of PGD and PGS, and it is to be expected that whole-genome sequencing will also follow. Current ethical and empirical sociological research on embryo testing focuses on PGD as it is practiced now. However, empirical research and systematic reflection regarding the impact of comprehensive techniques for embryo testing is missing. In order to understand the potential of this technology and to be able to adequately foresee its implications, we held an expert panel with seven pioneers in PGD. We conducted an expert panel in October 2011 with seven PGD pioneers from Belgium, The Netherlands, Germany and the UK. Participants expected the use of comprehensive techniques in the context of PGD. However, the introduction of these techniques in embryo testing requires timely ethical reflection as it involves a shift from choosing an embryo without a particular genetic disease (i.e. PGD) or most likely to result in a successful pregnancy (i.e. PGS) to choosing the best embryo based on a much wider set of criteria. Such ethical reflection should take account of current technical and biological limitations and also of current uncertainties with regard to the meaning of genetic variance. However, ethicists should also not be afraid to look into the future. There was a general agreement that embryo testing will be increasingly preceded by comprehensive preconception screening, thus enabling smart combinations of genetic testing. The group was composed of seven participants from

Medium scale fire tests of propane tanks to study the boiling liquid expanding vapour explosion (BLEVE) and transient two-phase jet release

Energy Technology Data Exchange (ETDEWEB)

Ye Zhifei

1994-07-01

A series of medium scale fire tests were conducted to study boiling liquid expanding vapour explosions (BLEVE) and transient jet releases resulting from thermally induced propane tank ruptures. The tests were conducted using commercial propane contained in automotive propane tanks with a capacity of ca 400 liters. The tanks were brought to failure using a combination of torch and pool fire impingement. Instrumentation was included to measure internal pressure, liquid, vapour and wall temperature distribution, tank and lading mass, external blast overpressure, and fireball thermal radiation. Video and still cameras were used to record the fireball and jet fire shapes and dimensions. Two different kinds of BLEVE failure were observed. For very weak tanks the BLEVE was a single step process where the rupture propagated rapidly along the length of the tank. The duration of these events was measured in milliseconds and it is suggested that the process is driven by the vapour space energy. The other type of BLEVE was a two step process where a crack would start in a weakened area, arrest in a stronger part of the tank, and then start again to end in catastrophic failure. Initial failure and jet type release results in violent boiling and pressure recovery in the tank, leading to restart of the crack and catastrophic failure. Time duration is measured in seconds, and is driven by energy stored in the liquid. A computer model was developed to simulate the transient jet release resulting from finite tank failures, and can predict transient mass flow, tank pressure decay, visible flame length and jet fire thermal radiation. 253 refs., 132 figs., 29 tabs.

Evaluating the Navy’s Enlisted Accessions Testing Program Based on Future Talent Needs

Science.gov (United States)

2017-03-01

NAVY’S ENLISTED ACCESSIONS TESTING PROGRAM BASED ON FUTURE TALENT NEEDS by Shereka F. Riley March 2017 Thesis Advisor: Joseph Sullivan Co...REPORT TYPE AND DATES COVERED Master’s thesis 4. TITLE AND SUBTITLE EVALUATING THE NAVY’S ENLISTED ACCESSIONS TESTING PROGRAM BASED ON FUTURE TALENT ...public release. Distribution is unlimited. 12b. DISTRIBUTION CODE 13. ABSTRACT (maximum 200 words) In recent years, non-defense related industries

Structural evaluation of electrosleeved tubes under severe accident transients

International Nuclear Information System (INIS)

Majumdar, S.

1999-01-01

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

Vertical axis wind turbine drive train transient dynamics

Science.gov (United States)

Clauss, D. B.; Carne, T. G.

1982-01-01

Start up of a vertical axis wind turbine causes transient torque oscillations in the drive train with peak torques which may be over two and one half times the rated torque of the turbine. A computer code, based on a lumped parameter model of the drive train, was developed and tested for the low cost 17 meter turbine; the results show excellent agreement with field data. The code was used to predict the effect of a slip clutch on transient torque oscillations. It was demonstrated that a slip clutch located between the motor and brake can reduce peak torques by thirty eight percent.

Spherical Indentation Techniques for Creep Property Evaluation Considering Transient Creep

Energy Technology Data Exchange (ETDEWEB)

Lim, Dongkyu; Kim, Minsoo; Lee, Hyungyil [Sogang Univ., Seoul, (Korea, Republic of); Lee, Jin Haeng [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-11-15

Creep through nanoindentations has attracted increasing research attention in recent years. Many studies related to indentation creep tests, however, have simply focused on the characteristics of steady-state creep, and there exist wide discrepancies between the uniaxial test and the indentation test. In this study, we performed a computational simulation of spherical indentations, and we proposed a method for evaluating the creep properties onsidering transient creep. We investigated the material behavior with variation of creep properties and expressed it using regression equations for normalized variables. We finally developed a program to evaluate the creep properties considering transient creep. By using the proposed method, we successfully obtained creep exponents with an average error less than 1.1 and creep coefficients with an average error less than 2.3 from the load-depth curve.

Spherical Indentation Techniques for Creep Property Evaluation Considering Transient Creep

International Nuclear Information System (INIS)

Lim, Dongkyu; Kim, Minsoo; Lee, Hyungyil; Lee, Jin Haeng

2013-01-01

Creep through nanoindentations has attracted increasing research attention in recent years. Many studies related to indentation creep tests, however, have simply focused on the characteristics of steady-state creep, and there exist wide discrepancies between the uniaxial test and the indentation test. In this study, we performed a computational simulation of spherical indentations, and we proposed a method for evaluating the creep properties onsidering transient creep. We investigated the material behavior with variation of creep properties and expressed it using regression equations for normalized variables. We finally developed a program to evaluate the creep properties considering transient creep. By using the proposed method, we successfully obtained creep exponents with an average error less than 1.1 and creep coefficients with an average error less than 2.3 from the load-depth curve

Stress state dependence of transient irradiation creep in 20% cold worked 316 stainless steel

International Nuclear Information System (INIS)

Foster, J.P.; Gilbert, E.R.

1998-01-01

Irradiation creep tests were performed in fast reactors using the stress states of uniaxial tension, biaxial tension, bending and torsion. In order to compare the saturated transient strain irradiation creep component, the test data were converted to equivalent strain and equivalent stress. The saturated transient irradiation creep component was observed to depend on the stress state. The highest value was exhibited by the uniaxial tension stress state, and the lowest by the torsion stress state. The biaxial tension and bending stress state transient component values were intermediate. This behavior appears to be related to the dislocation or microscopic substructure resulting from fabrication processing and the applied stress direction. (orig.)

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

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The first OECD LOFT experiment was conducted on February 20, 1983. It was designed to evaluate the generic PWR system response during a complete loss-of-feedwater transient. The objective of the experiment was to investigate the performance of primary 'feed and bleed' using a 'bleed' from the PORV and 'feed' from the HPIS to provide decay heat removal and system pressure reduction while maintaining the primary coolant inventory. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

Transient waveform acquisition system for the ELMO Bumpy Torus

International Nuclear Information System (INIS)

Young, K.G.; Burris, R.D.; Hillis, D.H.; Overbey, D.R.

1984-10-01

The transient waveform system described in this report is designed to acquire analog waveforms from the ELMO Bumpy Torus (EBT) diagnostic experiments. Pressure, density, synchrotron radiation, etc., are acquired and digitized with a Kinetic Systems TR812 transient recorder and associated modules located in a CAMAC crate. The system can simultaneously acquire, display, and transmit sets of data consisting of identification parameters and up to 1024 data points for 1 to 64 input signals (frequency range = 0.01 pulse/s to 100 kHz) of data every one or more minutes; thus, it can run continuously without operator intervention. The data are taken on a VAX 11/780 and transmitted to a data base on a DECSystem-10. To aid the programmer in making future modifications to the system, detailed documentation using the Yourdon structural methods has been given

An overview of FFTF [Fast Flux Test Facility] contributions to Liquid Metal Reactor Safety

International Nuclear Information System (INIS)

Waltar, A.E.; Padilla, A. Jr.

1990-11-01

The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs

On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

International Nuclear Information System (INIS)

Freels, J.D.

1993-01-01

This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed

Transient heat transfer for forced convection flow of helium gas

International Nuclear Information System (INIS)

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

1999-01-01

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

Efficacy and Tolerability of Indiplon in Transient Insomnia

Science.gov (United States)

Rosenberg, Russell; Roth, Thomas; Scharf, Martin B.; Lankford, D. Alan; Farber, Robert

2007-01-01

Objectives: The efficacy of indiplon was evaluated by polysomnography (PSG) in an experimental model of transient insomnia consisting of the first night effect combined with a 2-hour phase advance. Methods: Healthy volunteers age 21–64 years (N=593; 62% female; mean (± SEM) years, 32±0.39) were randomized to double-blind treatment with a single nighttime dose of indiplon (10 mg or 20 mg) or placebo. PSG assessments included latency to persistent sleep (LPS, primary endpoint) and total sleep time (TST); self-report assessments included sleep quality (SQ); next day residual effects were evaluated by the Digit Symbol Substitution Test (DSST), Symbol Copying Test (SCT), and a Visual Analog Scale of sleepiness (VAS). Results: LPS mean (± SEM) values were significantly reduced on indiplon 10 mg (21.2±1.5 minutes) and indiplon 20 mg (16.8±1.1 minutes) compared to placebo (33.1±2.5minutes; p inducing sleep, increasing sleep duration, and improving overall sleep quality without next day residual effects in healthy volunteers in a model of transient insomnia. Citation: Rosenberg R; Roth T; Scharf MB et al. Efficacy and tolerability of indiplon in transient insomnia. J Clin Sleep Med 2007;3(4):374-379. PMID:17694726

Crack propagation tests on the fundamental structure under cyclic thermal transients. Results of nondestructive inspection for cracks

International Nuclear Information System (INIS)

Kobayashi, S.; Horikiri, M.

2001-06-01

This report shows the results of crack inspection in crack propagation tests that were carried out at the Air-cooling Thermal Transient Test Facility (ATTF). Test specimens were made of 304 type austenitic stainless steel, and they were the same cylindrical shape, 1,500 mm in height, 130 mm in outer diameter and 30 mm in thickness. And they had initial slits machined on inner surfaces. Firstly the specimens were heated up to 650degC in a furnace, then cooled by pressurized air blowing through the specimen for 90 seconds. These cyclic changes of temperature gradients in the wall of specimens were loaded. Specimens were tested for several years. The specimen No. CPTT-102 with machined two circumferential slits and two semi-elliptical slits was tested up to 10,000 cycles. And the specimen No. CPTT-103 with machined six semi-elliptical slits of different length respectively was tested up to 5,000 cycles. Cracks of specimens were inspected nondestructively for a giving cycle in these tests. Applied inspection methods were ultra-sonic testing, potential-drop method and inner surface observation. Ultra-sonic testing was carried out by applying the pulse-echo method. Potential-drop testing was carried out by measurement of localized constant direct current beyond cracks. Photographs of the inner surface of specimens were taken using a bore-scope. The results of ultra-sonic testing have been close to destructive test results. The depth of crack by the potential-drop method was almost corresponding to destructive test results, too. Photographs of the inner surface were synthesized by the computer, and connection between main crack and hair crack was observed. (author)

Effect of additional holes on transient thermal fatigue life of gas turbine casing

Directory of Open Access Journals (Sweden)

H. Bazvandi

2017-10-01

Full Text Available Gas turbines casings are susceptible to cracking at the edge of eccentric pin hole, which is the most likely position for crack initiation and propagation. This paper describes the improvement of transient thermal fatigue crack propagation life of gas turbines casings through the application of additional holes. The crack position and direction was determined using non-destructive tests. A series of finite element patterns were developed and tested in ASTM-A395 elastic perfectly-plastic ductile cast iron. The effect of arrangement of additional holes on transient thermal fatigue behavior of gas turbines casings containing hole edge cracks was investigated. ABAQUS finite element package and Zencrack fracture mechanics code were used for modeling. The effect of the reduction of transient thermal stress distribution around the eccentric pin hole on the transient thermal fatigue crack propagation life of the gas turbines casings was discussed. The result shows that transient thermal fatigue crack propagation life could be extended by applying additional holes of larger diameter and decreased by increasing the vertical distance, angle, and distance between the eccentric pin hole and the additional holes. The results from the numerical predictions were compared with experimental data.

Prenatal Diagnosis of Transient Abnormal Myelopoiesis in a Down Syndrome Fetus

International Nuclear Information System (INIS)

Kim, Gwang Jun; Lee, Eun Sil

2009-01-01

We report a case of transient abnormal myelopoiesis in a Down syndrome fetus diagnosed at 28 +3 weeks of gestation that rapidly progressed to intrauterine death 10 days later. Fetal hepatosplenomegaly with cerebral ventriculomegaly, although not specific, may be a suggestive finding of Down syndrome with transient abnormal myelopoiesis. Prompt fetal blood sampling for liver function test and chromosomal analysis are mandatory for early detection and management

Analysis of the Mannshan Unit 2 full load rejection transient

International Nuclear Information System (INIS)

Kang, J.C.; Pei, B.S.; Yu, G.P.; Yuann, R.Y.

1987-01-01

Mannshan Unit 2 is a Westinghouse three-loop pressurized water reactor with a rated core power of 2775 MW(thermal) and a rated core flow of 4702 kg/s. Before full power operation, a planned net load rejection was performed during the startup test by opening the main transformer highside breakers. The generator power rapidly reduced to station load. All 16 steam dump valves immediately popped open, and control bank-D rods automatically stepped in as the temperature difference T/sub avg/ - T/sub ref/ reached a programmed 2.8 0 C. Nuclear power decreased smoothly as control rods were inserted into the core. The pressurizer pressure and liquid levels also dropped. Neither safety injection nor reactor trip occurred during this transient. The test was done to verify that the whole system would function properly under a transient to keep the reactor from scramming and that the vessel integrity would also be protected. In this study, which is the preliminary stage of RELAP5/MOD2 transient simulation of the Mannshan PWR plants, system thermal-hydraulic response is tested first and isolated from the neutronic effects. The variation of core power versus time curve was extracted from the power test data to serve as a time varying boundary condition. The comparison of the analytical results of four major parameters (pressurizer pressure, average temperature of the core, steam dump flow rate, and feedwater flow rate) from RELAP5/MOD2 and the power test data is illustrated

Power Testing in Basketball: Current Practice and Future Recommendations.

Science.gov (United States)

Wen, Neal; Dalbo, Vincent J; Burgos, Bill; Pyne, David B; Scanlan, Aaron T

2018-02-01

Numerous foundational movements performed during basketball are predicated on underlying power-related attributes, including speed, change-of-direction, and jumping. Accordingly, fitness testing batteries for basketball have incorporated an assortment of linear speed tests, change-of-direction tests, and jump tests. However, due to the wide variety of testing options it is difficult for basketball practitioners to select appropriate testing protocols for the assessment of power-related attributes. As a result, there is a need to review the relevant literature to identify game-specific, power-related attributes important in basketball and the most appropriate tests available to assess power-related attributes for basketball practitioners. Therefore, the aims of this review were to: (1) identify essential power-related attributes important in basketball; (2) discuss the suitability of common and novel power-related tests; and (3) provide recommendations for future research and best practice approaches for basketball coaching staff. In this review, we propose a series of novel tests that are more targeted and specific to basketball movements including: (1) 5-m and 10-m linear sprints, (2) Modified Agility T-Test, (3) Change-of-Direction Deficit, (4) lateral bound, (5) Sargent jump, (6) one-step jump, and (7) isometric midthigh pull test. Improved testing of power-related attributes should enable basketball practitioners to develop targeted training plans for enhancing player performance.

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

International Nuclear Information System (INIS)

Mitry, A.M.

1985-01-01

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

Recent irradiation tests for future nuclear system at HANARO

Energy Technology Data Exchange (ETDEWEB)

Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-05-15

The capsule at HANARO is a device that evaluates the irradiation effects of nuclear materials and fuels, which can reproduce the environment of nuclear power plants and accelerate to reach to the end of life condition. As the integrity assessment and the extension of lifetime of nuclear power plants are recently considered as important issues in Korea, the requirements for irradiation test are gradually being increased. The capacity and capability irradiation tests at HANARO are becoming important because Korea strives to develop SFR (Sodium-cooled Fast Reactor) and VHTR (Very High Temperature Reactor) among the future nuclear system and to export the research reactors and to develop the fusion reactor technology.

Fast Flux Test Facility: first three years of operation and future mission

International Nuclear Information System (INIS)

Peckinpaugh, C.L.; Newland, D.J.; Evans, E.A.

1985-03-01

In summary, the FFTF has proven to be a high performance, versatile test reactor. Results obtained during its first three years of operation - and those to be obtained in the coming years - are building a technology and experience base that is invaluable to future LMRs. The FFTF demonstrates proven LMR technology with a focus for the future and provides the US with international LMR technology leadership

PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

International Nuclear Information System (INIS)

Kirk, W.L.

1976-08-01

The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters

PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

Energy Technology Data Exchange (ETDEWEB)

Kirk, W.L. (comp.)

1976-08-01

The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

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)

Prenatal Diagnosis of Transient Abnormal Myelopoiesis in a Down Syndrome Fetus

Energy Technology Data Exchange (ETDEWEB)

Kim, Gwang Jun; Lee, Eun Sil [Chung-Ang University School of Medicine, Seoul (Korea, Republic of)

2009-04-15

We report a case of transient abnormal myelopoiesis in a Down syndrome fetus diagnosed at 28{sup +3} weeks of gestation that rapidly progressed to intrauterine death 10 days later. Fetal hepatosplenomegaly with cerebral ventriculomegaly, although not specific, may be a suggestive finding of Down syndrome with transient abnormal myelopoiesis. Prompt fetal blood sampling for liver function test and chromosomal analysis are mandatory for early detection and management.

Mercury flow tests (first report). Wall friction factor measurement tests and future tests plan

International Nuclear Information System (INIS)

Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro; Sudo, Yukio

1999-07-01

In the neutron science project at JAERI, we plan to inject a pulsed proton beam of a maximum power of 5 MW from a high intense proton accelerator into a mercury target in order to produce high energy neutrons of a magnitude of ten times or more than existing facilities. The neutrons produced by the facility will be utilized for advanced field of science such as the life sciences etc. An urgent issue in order to accomplish this project is the establishment of mercury target technology. With this in mind, a mercury experimental loop with the capacity to circulate mercury up to 15 L/min was constructed to perform thermal hydraulic tests, component tests and erosion characteristic tests. A measurement of the wall friction factor was carried out as a first step of the mercury flow tests, while testing the characteristic of components installed in the mercury loop. This report presents an outline of the mercury loop and experimental results of the wall friction factor measurement. From the wall friction factor measurement, it was made clear that the wettability of the mercury was improved with an increase of the loop operation time and at the same time the wall friction factors were increased. The measured wall friction factors were much lower than the values calculated by the Blasius equation at the beginning of the loop operation because of wall slip caused by a non-wetted condition. They agreed well with the values calculated by the Blasius equation within a deviation of 10% when the sum of the operation time increased more than 11 hours. This report also introduces technical problems with a mercury circulation and future tests plan indispensable for the development of the mercury target. (author)

412th Test Engineering Group Vision for Future Knowledge Management (KM)

Science.gov (United States)

2018-05-17

Presentation 3. DATES COVERED (From - To) 17 May 2018 4. TITLE AND SUBTITLE 412th Test Engineering Group Vision for Future Knowledge Management (KM... Engineering Group 307 E. Popson Ave Edwards AFB, CA 93523 8. PERFORMING ORGANIZATION REPORT NUMBER 412TW-PA...centers for the TENG test customers to allow the data to be readily available within minutes of a flight, for the data to be organized so that the engineer

Building blocks for future detectors: Silicon test masses and 1550 nm laser light

International Nuclear Information System (INIS)

Schnabel, R; Britzger, M; Burmeister, O; Danzmann, K; Duck, J; Eberle, T; Friedrich, D; Luck, H; Mehmet, M; Steinlechner, S; Willke, B; Brueckner, F; Nawrodt, R

2010-01-01

Current interferometric gravitational wave detectors use the combination of quasi-monochromatic, continuous-wave laser light at 1064 nm and fused silica test masses at room temperature. Detectors of the third generation, such as the Einstein-Telescope, will involve a considerable sensitivity increase. The combination of 1550 nm laser radiation and crystalline silicon test masses at low temperatures might be important ingredients in order to achieve the sensitivity goal. Here we compare some properties of the fused silica and silicon test mass materials relevant for decreasing the thermal noise in future detectors as well as the recent technology achievements in the preparation of laser radiation at 1064 nm and 1550 nm relevant for decreasing the quantum noise. We conclude that silicon test masses and 1550 nm laser light have the potential to form the future building blocks of gravitational wave detection.

Investigation of practical use situation and performance for electric transient analysis programs in the U.S. nuclear power plants

International Nuclear Information System (INIS)

Shimada, Yoshio

2010-01-01

The purposes of the present study are firstly to investigate the status of practical use of electric transient analysis programs used in U.S. nuclear power plants, which has been extracted as good examples from the information analysis of overseas troubles, and secondly to select a program to be recommended for use in implementing electric transient analysis in domestic nuclear power plants. In addition, to promote its practical use, a selected electric transient analysis program was tested by simulating the transient response during a load sequence test of an emergency diesel generator (EDG) in a domestic representative nuclear plant to evaluate its simulation accuracy by comparing its result with the measured plant data. The results obtained are as follows: (1) In U.S. nuclear power plants, simulations using electric transient analysis programs, such as ETAP, EMPT, etc., are widely performed, which contributed to improve the plant safety. (2) A selected transient analysis program EMTP was verified in its accuracy in terms of transient response of active power, current, voltage and frequency of the EDG during the load sequence test in a domestic representative nuclear power plant. (author)

Investigation of Thermal and Vacuum Transients on the LHC Prototype Magnet String

CERN Document Server

Cruikshank, P; Riddone, G; Tavian, L

1996-01-01

The prototype magnet string, described in a companion paper, is a full-scale working model of a 50-m length of the future Large Hadron Collider (LHC), CERN's new accelerator project, which will use high-field superconducting magnets operating below 2 K in superfluid helium. As such, it provides an excellent test bed for practising standard operating modes of LHC insulation vacuum and cryogenics, as well as for experimentally assessing accidental behaviour and failure modes, and thus verifying design calculations. We present experimental investigation of insulation vacuum pumpdown, magnet forced-flow cooldown and warmup, and evolution of residual vacuum pressures and temperatures in natural warmup, as well as catastrophic loss of insulation vacuum. In all these transient modes, experimental results are compared with simulated behaviour, using a non-linear, one-dimensional thermal model of the magnet string.

Effects of transient conditions on exhaust emissions from two non-road diesel engines

International Nuclear Information System (INIS)

Lindgren, M.; Hansson, P.-A.

2004-01-01

Growing interest in quantifying and reducing the amount of engine emissions of carbon monoxide, hydrocarbons, and nitrogen oxides loading the environment has led to increasingly tighter environmental regulations. However, current non-road emission standards are performed according to a steady-state test cycle, which does not include transient effects and thus underestimates the amount of emissions produced in real use of the engine. This study quantifies the effects of transients in engine speed and torque on the fuel consumption and emissions from two diesel engines intended for non-road mobile machinery. Fuel consumption and emissions from the engines were measured in an engine dynamometer during various transient load conditions. The results showed that during fast transients, the measured fuel consumption was up to twice as high as the corresponding steady-state load conditions. The effects of transients on emissions of nitrogen oxides were even greater, as were the effects of transient load increase with increasing transient conditions i. e. rate of change. The results showed that the effect of transients on fuel consumption and emissions were also dependent on the type of diesel injection pump and the engine equipment used. Furthermore, the results indicated that the air/fuel ratio was an important contributor to the emission formation process during transient loads. (Author)

Analysis of the transient compressible vapor flow in heat pipes

Science.gov (United States)

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

1989-01-01

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

Analysis of the transient compressible vapor flow in heat pipe

International Nuclear Information System (INIS)

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

1989-07-01

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

Analysis of the transient compressible vapor flow in heat pipe

Science.gov (United States)

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

1989-01-01

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

Transient failure behavior of HT9

International Nuclear Information System (INIS)

Huang, F.H.

1994-07-01

Alloy HT9 has-been chosen as candidate materials for fast and fusion reactor applications because the.material exhibits excellent resistance to void swelling. However, ferritic alloys are known to undergo a ductile-brittle transition as the test temperature is decreased. This inherent problem has limited their applications to reactor component materials subjected to low neutron exposures. Despite the ductile-brittle transition problem, results show that the materials exhibit superior resistance to fracture under very high neutron fluences at irradiation temperatures above 380C. Results also show that the transient behavior for HT9 cladding specimens taken from the fuel column region and cladding taken from outside the fuel column or unirradiated cladding are the same. HT9 cladding maintained its transient strength with irradiation to a fluence of 9 x 10 22 n/cm 2 (E > 0.1 MeV)

The fuel performance code future

International Nuclear Information System (INIS)

Ronchi, C.; Van de Laar, J.

1988-01-01

The paper describes the LWR version of the fuel performance code FUTURE, which was recently developed to calculate the fuel response (swelling, cladding deformation, release) to reactor transient conditions, starting from a broad-based description of the processes of major concern. The main physical models assumed are presented together with the scheme of the computer program

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

Can We Remember Future Actions yet Forget the Last Two Minutes? Study in Transient Global Amnesia

Science.gov (United States)

Hainselin, Mathieu; Quinette, Peggy; Desgranges, Beatrice; Martinaud, Olivier; Hannequin, Didier; de La Sayette, Vincent; Viader, Fausto; Eustache, Francis

2011-01-01

Transient global amnesia (TGA) is a clinical syndrome characterized by the abrupt onset of a massive episodic memory deficit that spares other cognitive functions. If the anterograde dimension is known to be impaired in TGA, researchers have yet to investigate prospective memory (PM)--which involves remembering to perform an intended action at…

A transient search using combined human and machine classifications

Science.gov (United States)

Wright, Darryl E.; Lintott, Chris J.; Smartt, Stephen J.; Smith, Ken W.; Fortson, Lucy; Trouille, Laura; Allen, Campbell R.; Beck, Melanie; Bouslog, Mark C.; Boyer, Amy; Chambers, K. C.; Flewelling, Heather; Granger, Will; Magnier, Eugene A.; McMaster, Adam; Miller, Grant R. M.; O'Donnell, James E.; Simmons, Brooke; Spiers, Helen; Tonry, John L.; Veldthuis, Marten; Wainscoat, Richard J.; Waters, Chris; Willman, Mark; Wolfenbarger, Zach; Young, Dave R.

2017-12-01

Large modern surveys require efficient review of data in order to find transient sources such as supernovae, and to distinguish such sources from artefacts and noise. Much effort has been put into the development of automatic algorithms, but surveys still rely on human review of targets. This paper presents an integrated system for the identification of supernovae in data from Pan-STARRS1, combining classifications from volunteers participating in a citizen science project with those from a convolutional neural network. The unique aspect of this work is the deployment, in combination, of both human and machine classifications for near real-time discovery in an astronomical project. We show that the combination of the two methods outperforms either one used individually. This result has important implications for the future development of transient searches, especially in the era of Large Synoptic Survey Telescope and other large-throughput surveys.

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

Transient diagnosis system using quantum-inspired computing and Minkowski distance

Energy Technology Data Exchange (ETDEWEB)

Nicolau, Andressa dos Santos; Schirru, Roberto, E-mail: andressa@lmp.ufrj.b, E-mail: schirru@lmp.ufrj.b [Federal University of Rio de Janeiro (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program

2011-07-01

This paper proposes a diagnosis system model for identification of transient in a PWR nuclear power plant, optimized by the Quantum Inspired Evolutionary Algorithm - QEA in order to help nuclear power plant operator reduce his cognitive load and increase his available time to maintain the plant operating in a safe condition. This method was developed in order to be able to recognize the normal condition and three accidents of the design basis list of the nuclear power plant Angra 2, postulated in the Final Safety Analysis Report (FSAR). This System compares the similarly distance between the set of variables of the anomalous event, in a given time t, and the centroids of the design-basis transient variables. The lower similarly distance indicates the class of the transient to which the anomalous event belongs. The QEA was then used to find the best position of the centroids of each class of the selected transients. Such positions maximize the number of the correct classifications. Unlike the diagnosis system proposed in the literature, Minkowski distance was employed to calculate the similarity distance. The signatures of four transients were submitted to 1% and 2% of noise, and tested with prototype vector found by QEA. The results showed that the present transient diagnostic system was successfully implemented in the nuclear accident identification problem and was compatible with the techniques presented in the literature. (author)

Transient diagnosis system using quantum-inspired computing and Minkowski distance

International Nuclear Information System (INIS)

Nicolau, Andressa dos Santos; Schirru, Roberto

2011-01-01

This paper proposes a diagnosis system model for identification of transient in a PWR nuclear power plant, optimized by the Quantum Inspired Evolutionary Algorithm - QEA in order to help nuclear power plant operator reduce his cognitive load and increase his available time to maintain the plant operating in a safe condition. This method was developed in order to be able to recognize the normal condition and three accidents of the design basis list of the nuclear power plant Angra 2, postulated in the Final Safety Analysis Report (FSAR). This System compares the similarly distance between the set of variables of the anomalous event, in a given time t, and the centroids of the design-basis transient variables. The lower similarly distance indicates the class of the transient to which the anomalous event belongs. The QEA was then used to find the best position of the centroids of each class of the selected transients. Such positions maximize the number of the correct classifications. Unlike the diagnosis system proposed in the literature, Minkowski distance was employed to calculate the similarity distance. The signatures of four transients were submitted to 1% and 2% of noise, and tested with prototype vector found by QEA. The results showed that the present transient diagnostic system was successfully implemented in the nuclear accident identification problem and was compatible with the techniques presented in the literature. (author)

Heavy Ion Microbeam- and Broadbeam-Induced Current Transients in SiGe HBTs

Science.gov (United States)

Pellish, Jonathan A.; Reed, R. A.; McMorrow, D.; Vizkelethy, G.; Ferlet-Cavrois, V.; Baggio, J.; Duhamel, O.; Moen, K. A.; Phillips, S. D.; Diestelhorst, R. M.;

Transient hypothyroidism in the newborn: to treat or not to treat

Science.gov (United States)

Kanike, Neelakanta; Davis, Ajuah

2017-01-01

Transient congenital hypothyroidism (CH) refers to a temporary deficiency of thyroid hormone identified after birth, with low thyroxine (T4) and elevated thyrotropin (TSH), which later recovers to improved thyroxine production, typically in first few months of infancy. Approximately 17% to 40% of children diagnosed with CH by newborn screening (NBS) programs were later determined to have transient hypothyroidism. Causes of transient CH are prematurity, iodine deficiency, maternal thyrotropin receptor blocking antibodies, maternal intake of anti-thyroid drugs, maternal or neonatal iodine exposure, loss of function mutations and hepatic hemangiomas. The classic clinical symptoms and signs of CH are usually absent immediately after birth in vast majority of infants due to temporary protection from maternal thyroxine. NBS has been largely successful in preventing intellectual disability by early detection of CH by performing thyroid function tests in infants with abnormal screening results. In this review we present the evidence for decision making regarding treatment vs. withholding treatment in infants with transient CH and present a rational approach to identifying transient CH based on American Academy of Pediatrics (AAP) recommendation. PMID:29184815

Tests of the SNR fuel pin behaviour in case of operational transients in the HFR Petten

International Nuclear Information System (INIS)

Plitz, H.

1989-05-01

The loadings on fast reactor fuel pins under operational transients (power and temperature increases in the design area) have been studied in the High-Flux-Reactor HFR in Petten with sodium cooled irradiation capsules. The results of the first campaign of transient experiments are described in the report. No cladding defects have been observed, and the fuel pins of the Mark-I and Mark-II type resisted to linear power levels of more than 800 W/cm, thus demonstrating the required design margins. The plans for further experiments are outlined

Analysis of Peach Bottom turbine trip tests

International Nuclear Information System (INIS)

Cheng, H.S.; Lu, M.S.; Hsu, C.J.; Shier, W.G.; Diamond, D.J.; Levine, M.M.; Odar, F.

1979-01-01

Current interest in the analysis of turbine trip transients has been generated by the recent tests performed at the Peach Bottom (Unit 2) reactor. Three tests, simulating turbine trip transients, were performed at different initial power and coolant flow conditions. The data from these tests provide considerable information to aid qualification of computer codes that are currently used in BWR design analysis. The results are presented of an analysis of a turbine trip transient using the RELAP-3B and the BNL-TWIGL computer codes. Specific results are provided comparing the calculated reactor power and system pressures with the test data. Excellent agreement for all three test transients is evident from the comparisons

Study of transient burnout characteristics under flow reduction condition

International Nuclear Information System (INIS)

Iwamura, Takamichi; Kuroyanagi, Toshiyuki

1984-03-01

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

A study of the transient performance of annular hydrostatic journal bearings in liquid oxygen

Science.gov (United States)

Scharrer, J. K.; Tellier, J. G.; Hibbs, R. I.

1992-07-01

A test apparatus was used to simulate a cryogenic turbopump start transient in order to determine the liftoff and touchdown speed and amount of wear of an annular hydrostatic bearing in liquid oxygen. The bearing was made of sterling silver and the journal made of Inconel 718. The target application of this configuration is the pump end bearing of the Space Shuttle Main Engine High Pressure Liquid Oxygen Turbopump. Sixty-one transient cycles were performed in liquid oxygen with an additional three tests in liquid nitrogen to certify the test facility and configuration. The bearing showed no appreciable wear during the testing, and the results indicate that the performance of the bearing was not significantly degraded during the testing.

Transient and steady-state selection in the striatal microcircuit

Directory of Open Access Journals (Sweden)

Adam eTomkins

2014-01-01

Full Text Available Although the basal ganglia have been widely studied and implicated in signal processing and action selection, little information is known about the active role the striatal microcircuit plays in action selection in the basal ganglia-thalamo-cortical loops. To address this knowledge gap we use a large scale three dimensional spiking model of the striatum, combined with a rate coded model of the basal ganglia-thalamo-cortical loop, to asses the computational role the striatum plays in action selection. We identify a robust transient phenomena generated by the striatal microcircuit, which temporarily enhances the difference between two competing cortical inputs. We show that this transient is sufficient to modulate decision making in the basal ganglia-thalamo-cortical circuit. We also find that the transient selection originates from a novel adaptation effect in single striatal projection neurons, which is amenable to experimental testing. Finally, we compared transient selection with models implementing classical steady-state selection. We challenged both forms of model to account for recent reports of paradoxically enhanced response selection in Huntington's Disease patients. We found that steady-state selection was uniformly impaired under all simulated Huntington's conditions, but transient selection was enhanced given a sufficient Huntington's-like increase in NMDA receptor sensitivity. Thus our models provide an intriguing hypothesis for the mechanisms underlying the paradoxical cognitive improvements in manifest Huntington's patients.

Psychiatric genetic testing: Attitudes and intentions among future users and providers

DEFF Research Database (Denmark)

Laegsgaard, Mett Marri; Mors, Ole

2008-01-01

as a guide in this field, but the optimal utilization of genetic testing has also been recognized to depend on knowledge of the potential consumers' attitudes. To provide knowledge to inform the public debate on mental illness and genetics, and the future conducting of psychiatric genetic testing....... Psychiatric and somatic genetic testing attracted the same amounts of accept. General attitudes toward access to psychiatric genetic testing and information revealed substantial support for bioethical principles of autonomy and privacy. However, questions describing more specific situations revealed......Psychiatric genetic research may eventually render possible psychiatric genetic testing. Whereas all genetic knowledge has certain characteristics raising ethical, legal, and social issues, psychiatric genetic knowledge adds more controversial issues. Ethical principles have been proposed...

Comprehensive cardiac rehabilitation for secondary prevention after transient ischemic attack or mild stroke: I: feasibility and risk factors.

Science.gov (United States)

Prior, Peter L; Hachinski, Vladimir; Unsworth, Karen; Chan, Richard; Mytka, Sharon; O'Callaghan, Christina; Suskin, Neville

2011-11-01

Comprehensive cardiac rehabilitation (CCR), which integrates structured lifestyle interventions and medications, reduces morbidity and mortality among cardiac patients. CCR has not typically been used with cerebrovascular populations, despite important commonalities with heart patients. We tested feasibility and effectiveness of 6-month outpatient CCR for secondary prevention after transient ischemic attack or mild, nondisabling stroke. This article presents risk factors. A future article will discuss psychological outcomes. Consecutive consenting subjects having sustained a transient ischemic attack or mild, nondisabling stroke within the previous 12 months (mean, 11.5 weeks; event-to-CCR entry) with ≥1 vascular risk factor, were recruited from a stroke prevention clinic providing usual care. We measured 6-month CCR outcomes following a prospective cohort design. Of 110 subjects recruited from January 2005 to April 2006, 100 subjects (mean age, 64.9 years; 46 women) entered and 80 subjects completed CCR. We obtained favorable, significant intake-to-exit changes in: aerobic capacity (+31.4%; Pstroke, offering a promising model for vascular protection across chronic disease entities. We know of no similar previous investigation, and are now conducting a randomized trial.

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.

A strategy to unveil transient sources of ultra-high-energy cosmic rays

Directory of Open Access Journals (Sweden)

Takami Hajime

2013-06-01

Full Text Available Transient generation of ultra-high-energy cosmic rays (UHECRs has been motivated from promising candidates of UHECR sources such as gamma-ray bursts, flares of active galactic nuclei, and newly born neutron stars and magnetars. Here we propose a strategy to unveil transient sources of UHECRs from UHECR experiments. We demonstrate that the rate of UHECR bursts and/or flares is related to the apparent number density of UHECR sources, which is the number density estimated on the assumption of steady sources, and the time-profile spread of the bursts produced by cosmic magnetic fields. The apparent number density strongly depends on UHECR energies under a given rate of the bursts, which becomes observational evidence of transient sources. It is saturated at the number density of host galaxies of UHECR sources. We also derive constraints on the UHECR burst rate and/or energy budget of UHECRs per source as a function of the apparent source number density by using models of cosmic magnetic fields. In order to obtain a precise constraint of the UHECR burst rate, high event statistics above ∼ 1020 eV for evaluating the apparent source number density at the highest energies and better knowledge on cosmic magnetic fields by future observations and/or simulations to better estimate the time-profile spread of UHECR bursts are required. The estimated rate allows us to constrain transient UHECR sources by being compared with the occurrence rates of known energetic transient phenomena.

International standard problem (ISP) No. 43 Rapid boron-dilution transient tests for code verification. Comparison report

International Nuclear Information System (INIS)

2001-03-01

International Standard Problem No. 43 (ISP 43) addresses the nuclear industries present capabilities of simulating fluid dynamics aspects of a subset of rapid boron dilution transients. Specifically, the exercise focuses on the sequence involving the transport of a boron-dilute slug through the actuation of a pump. The slug is formed on the primary side of the steam generator as a consequence of in interfacing system leak from the secondary un-borated coolant. Experimental data was collected using the University of Maryland 2 x 4 Thermalhydraulic Loop (UM 2 x 4 Loop) and the Boron-mixing Visualization Facility. Two blind test series were proposed during the first workshop (October 1998) and refined using participant input. The first series, test series A, deals with the injection of a front, i.e., a single interface between borated and dilute fluids. The second blind series, test series B, is the more realistic injection of a slug, i.e., a dilute fluid volume preceded and followed by the borated coolant of the primary system. Data are collected in the UM 2 x 4 Loop and refined details are obtained from the Visualization Facility, which represents a replica of the Loop.s vessel downcomer. In the Loop experimental program, the dilute volume is simulated by cold water and the borated primary coolant is simulated by hot water. The Visualization Facility uses dye to mark the diluted front or slug. The measured boundary conditions for both test series include the initial temperature of the primary system, the front/slug injection flowrate and temperature, and the pressure drop across the core. Temperature data is collected at 185 thermocouple positions in the downcomer and 38 positions in the lower plenum. The advancement of the front/slug through the system is monitored at discrete horizontal levels that contain the thermocouples. The performance of codes is measured relative to a set of figures of merit. During the first workshop, the principal figure of merit was

Modelling structural systems for transient response analysis

International Nuclear Information System (INIS)

Melosh, R.J.

1975-01-01

This paper introduces and reports success of a direct means of determining the time periods in which a structural system behaves as a linear system. Numerical results are based on post fracture transient analyses of simplified nuclear piping systems. Knowledge of the linear response ranges will lead to improved analysis-test correlation and more efficient analyses. It permits direct use of data from physical tests in analysis and simplication of the analytical model and interpretation of its behavior. The paper presents a procedure for deducing linearity based on transient responses. Given the forcing functions and responses of discrete points of the system at various times, the process produces evidence of linearity and quantifies an adequate set of equations of motion. Results of use of the process with linear and nonlinear analyses of piping systems with damping illustrate its success. Results cover the application to data from mathematical system responses. The process is successfull with mathematical models. In loading ranges in which all modes are excited, eight digit accuracy of predictions are obtained from the equations of motion deduced. Small changes (less than 0.01%) in the norm of the transfer matrices are produced by manipulation errors for linear systems yielding evidence that nonlinearity is easily distinguished. Significant changes (greater than five %) are coincident with relatively large norms of the equilibrium correction vector in nonlinear analyses. The paper shows that deducing linearity and, when admissible, quantifying linear equations of motion from transient response data for piping systems can be achieved with accuracy comparable to that of response data

Transient fuel melting

International Nuclear Information System (INIS)

Roche, L.; Schmitz, F.

1982-10-01

The observation of micrographic documents from fuel after a CABRI test leads to postulate a specific mode of transient fuel melting during a rapid nuclear power excursion. When reaching the melt threshold, the bands which are characteristic for the solid state are broken statistically over a macroscopic region. The time of maintaining the fuel at the critical enthalpy level between solid and liquid is too short to lead to a phase separation. A significant life-time (approximately 1 second) of this intermediate ''unsolide'' state would have consequences on the variation of physical properties linked to the phase transition solid/liquid: viscosity, specific volume and (for the irradiated fuel) fission gas release [fr

The walking-induced transient hack concept is valid & relies on a transient early-exercise hypoxemia.

Directory of Open Access Journals (Sweden)

Antoine Bruneau

Full Text Available Decreased arterial oxygen pressure obtained at peak exercise is strong evidence of walking-induced hypoxemia, assuming that the lower pressure occurs just before exercise is stopped. Using empirical predefined models and transcutaneous oximetry, we have shown that some patients reporting exercise intolerance show a minimal value at the onset of walking and a post-exercise overshoot. These changes are referred to as transcutaneous "walking-induced transient hacks".In 245 patients, walking-induced transcutaneous oxygen pressure changes in the chest were analyzed using observer-independent clustering techniques. Clustering classes were compared to the profile types previously proposed with the cross-correlation technique. The classifications of patients according to both approaches were compared using kappa statistics. In 10 patients showing a hack on transcutaneous oximetry, we analyzed the results of direct iterative arterial sampling recorded during a new walking treadmill test.Clustering analysis resulted in 4 classes that closely fit the 4 most frequently proposed empirical models (cross-correlation coefficients: 0.93 to 0.97. The kappa between the two classifications was 0.865. In 10 patients showing transcutaneous hacks, the minimal direct arterial oxygen pressure value occurred at exercise onset, and these patients exhibited a recovery overshoot reaching a maximum at two minutes of recovery, confirming the walking-induced transient hypoxemia.In patients reporting exercise intolerance, transcutaneous oximetry could help to detect walking-induced transient hypoxemia, while peak-exercise arterial oximetry might be normal.

Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

International Nuclear Information System (INIS)

Johnson, G.D.; Hunter, C.W.

1978-06-01

Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results

An approach to DNI transients characterization for system evaluation

Science.gov (United States)

Feldhoff, Jan Fabian; Hirsch, Tobias

2017-06-01

The direct normal irradiance (DNI) is of utmost importance for concentrated solar power (CSP) plants. For annual yield prediction, a steady-state heat balance is made for each hour of the year or for a smaller time period such as 15 min with the corresponding average DNI value. However, short term DNI variations by clouds are ignored by this approach. In consequence, there is no information on the transient behavior of the plant and the question remains how the plant is influenced by the DNI disturbance. The paper intends to start a discussion on DNI characterization and its application to CSP. An approach to categorize the DNI behavior from a transient system point of view is presented by using purpose-/system-specific filters. Resulting DNI disturbance classes are proposed to directly compare different sites and technologies. They can be useful for better yield analysis and better commercial project selection in the future. An example on a once-through direct steam generation plant is provided.

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.

Transient Dynamics Analysis of The Reachstacker Speader Based On ANSYS

Directory of Open Access Journals (Sweden)

Shu Yu Feng

2016-01-01

Full Text Available Reachstacker is an indispensable handling machinery, it will inevitably lead to unbalanced force at the job site. This paper does transient dynamics analysis for the spreader mechanism, which is one of the most significance key components. We get dynamic response of the spreader in lifting instant, results not only provide a reference for designers to understand the mechanical characteristics of spreader comprehensively, but also bedding for the future research.

RFI flagging implications for short-duration transients

Science.gov (United States)

Cendes, Y.; Prasad, P.; Rowlinson, A.; Wijers, R. A. M. J.; Swinbank, J. D.; Law, C. J.; van der Horst, A. J.; Carbone, D.; Broderick, J. W.; Staley, T. D.; Stewart, A. J.; Huizinga, F.; Molenaar, G.; Alexov, A.; Bell, M. E.; Coenen, T.; Corbel, S.; Eislöffel, J.; Fender, R.; Grießmeier, J.-M.; Jonker, P.; Kramer, M.; Kuniyoshi, M.; Pietka, M.; Stappers, B.; Wise, M.; Zarka, P.

2018-04-01

With their wide fields of view and often relatively long coverage of any position in the sky in imaging survey mode, modern radio telescopes provide a data stream that is naturally suited to searching for rare transients. However, Radio Frequency Interference (RFI) can show up in the data stream in similar ways to such transients, and thus the normal pre-treatment of filtering RFI (flagging) may also remove astrophysical transients from the data stream before imaging. In this paper we investigate how standard flagging affects the detectability of such transients by examining the case of transient detection in an observing mode used for Low Frequency Array (LOFAR; van Haarlem et al., 2013) surveys. We quantify the fluence range of transients that would be detected, and the reduction of their SNR due to partial flagging. We find that transients with a duration close to the integration sampling time, as well as bright transients with durations on the order of tens of seconds, are completely flagged. For longer transients on the order of several tens of seconds to minutes, the flagging effects are not as severe, although part of the signal is lost. For these transients, we present a modified flagging strategy which mitigates the effect of flagging on transient signals. We also present a script which uses the differences between the two strategies, and known differences between transient RFI and astrophysical transients, to notify the observer when a potential transient is in the data stream.

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

Science.gov (United States)

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

1991-01-01

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

Current trends and future development in pharmacologic stress testing

International Nuclear Information System (INIS)

Bae, Jin Ho; Lee, Jae Tae

2005-01-01

Pharmacologic stress testing for myocardial perfusion imaging is a widely used noninvasive method for the evaluation of known or suspected coronary artery disease. The use of exercise for cardiac stress has been practiced for over 60 years and clinicians are familiar with its using. However, there are inevitable situations in which exercise stress is inappropriate. A large number of patients with cardiac problems are unable to exercise to their full potential due to comorbidity such as osteoarthritis, vascular disease and pulmonary disease and a standard exercise stress test for myocardial perfusion imaging is suboptimal means for assessment of coronary artery disease. This problem has led to the development of the pharmacologic stress test and to a great increase in its popularity. All of the currently used pharmacologic agents have well-documented diagnostic value. This review deals the physiological actions, clinical protocols, safety, nuclear imaging applications of currently available stress agents and future development of new vasodilating agents

Transient receptor potential channel polymorphisms are associated with the somatosensory function in neuropathic pain patients.

Directory of Open Access Journals (Sweden)

Andreas Binder

Full Text Available Transient receptor potential channels are important mediators of thermal and mechanical stimuli and play an important role in neuropathic pain. The contribution of hereditary variants in the genes of transient receptor potential channels to neuropathic pain is unknown. We investigated the frequency of transient receptor potential ankyrin 1, transient receptor potential melastin 8 and transient receptor potential vanilloid 1 single nucleotide polymorphisms and their impact on somatosensory abnormalities in neuropathic pain patients. Within the German Research Network on Neuropathic Pain (Deutscher Forscbungsverbund Neuropathischer Schmerz 371 neuropathic pain patients were phenotypically characterized using standardized quantitative sensory testing. Pyrosequencing was employed to determine a total of eleven single nucleotide polymorphisms in transient receptor potential channel genes of the neuropathic pain patients and a cohort of 253 German healthy volunteers. Associations of quantitative sensory testing parameters and single nucleotide polymorphisms between and within groups and subgroups, based on sensory phenotypes, were analyzed. Single nucleotide polymorphisms frequencies did not differ between both the cohorts. However, in neuropathic pain patients transient receptor potential ankyrin 1 710G>A (rs920829, E179K was associated with the presence of paradoxical heat sensation (p = 0.03, and transient receptor potential vanilloid 1 1911A>G (rs8065080, I585V with cold hypoalgesia (p = 0.0035. Two main subgroups characterized by preserved (1 and impaired (2 sensory function were identified. In subgroup 1 transient receptor potential vanilloid 1 1911A>G led to significantly less heat hyperalgesia, pinprick hyperalgesia and mechanical hypaesthesia (p = 0.006, p = 0.005 and pG (rs222747, M315I to cold hypaesthesia (p = 0.002, but there was absence of associations in subgroup 2. In this study we found no evidence that genetic

Transient analysis of mercury experimental loop using the RELAP5 code. 3rd report. Transient analysis using mercury properties

International Nuclear Information System (INIS)

Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro

2000-02-01

In order to promote the Neutron Science Project of JAERI, the design of a 5MW-spallation target system is in progress with the purpose of producing a practical neutron application while at the same time adhering to the highest levels of safety. To establish the safety of the target system, it is important to understand the transient behaviors during anticipated operational events of the system, and to design the safety protection systems for the safe termination of the transients. This report presents the analytical results of transient behaviors in the mercury experimental loop using mercury properties. At first, the analytical pressure distributions were compared with experimental data measured with the mercury experimental loop. The modeling data were modified to reproduce the actual pressure distributions of the mercury experimental loop. Then a loss of forced convection and a loss of coolant accident were analyzed. In the case of the pump trip, the transient analysis was conducted using two types of mercury pumps, the mechanical type pump with moment of inertia, and the electrical-magnetic type pump without moment of inertia. The results show there was no clear difference in the two analyses, since the mercury had a large inertia, which was 13.5 times that of the water. Moreover, in the case of a pipe rupture at the pump exit, a moderate pressure decrease was confirmed when a small breakage area existed in which the coolant flowed out gradually. Based on these results, it was appeared that the transient fluctuation of pressure in the mercury loop would not become large and accidents would have to be detected by small fluctuations in pressure. Based on these analyses, we plan to conduct a simulation test to verify the RELAP5 code, and then the analysis of a full-scale mercury system will be performed. (author)

Investigate Transient Behaviours and Select Appropriate Fault Protection Solutions of Uni-grounded AC Microgrids

Directory of Open Access Journals (Sweden)

Duong Minh Bui

2016-03-01

Full Text Available Transient situations of a uni-grounded low-voltage AC microgrid are simulated in this paper, which include different fault tests and operation transition test between the grid-connected and islanded modes of the uni-grounded microgrid. Based on transient simulation results, available fault protection methods are proposed for the main and back-up protection of a uni-grounded AC microgrid. Main contributions of this paper are (i analysing transient responses of a typically uni-grounded lowvoltage AC microgrid from line-to-line, single line-to-ground, three-phase faults and a microgrid operation transition test; and (ii proposing available fault protection methods for uni-grounded AC microgrids, such as non-directional/directional overcurrent protection solutions, under/over voltage protection solutions, differential protection, voltage-restrained overcurrent protection, and other protection principles not based on high fault currents (e.g. total harmonic distortion detection of phase currents and voltages, or protection methods using symmetrical sequence components of current and voltage.

LLL transient-electromagnetics-measurement facility

International Nuclear Information System (INIS)

Deadrick, F.J.; Miller, E.K.; Hudson, H.G.

1975-01-01

The operation and hardware of the Lawrence Livermore Laboratory's transient-electromagnetics (EM)-measurement facility are described. The transient-EM range is useful for determining the time-domain transient responses of structures to incident EM pulses. To illustrate the accuracy and utility of the EM-measurement facility, actual experimental measurements are compared to numerically computed values

Transient Model of a 10 MW Supercritical CO{sub 2} Brayton Cycle for Light Water Reactors by using MARS Code

Energy Technology Data Exchange (ETDEWEB)

Park, Joo-Hyun; Park, Hyun Sun; Kim, Moo Hwan [POSTECH, Pohang (Korea, Republic of); Bae, Sung Won; Cha, Jae-Eun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this study, recuperation cycle was chosen as a reference loop design and the MARS code was chosen as the transient cycle analysis code. Cycle design condition is focus on operation point of the light-water reactor. Development of a transient model was performed for 10MW-electron SCO{sub 2} coupled with light water reactors. In order to perform transient analysis, cycle transient model was developed and steady-state run was performed and presented in the paper. In this study, the transient model of SCO{sub 2} recuperation Brayton cycle was developed and implemented in MARS to study the steady-state simulation. We performed nodalization of the transient model using MARS code and obtained steady-state results. This study is shown that the supercritical CO{sub 2} Brayton cycle can be used as a power conversion system for light water reactors. Future work will include transient analysis such as partial road operation, power swing, start-up, and shutdown. Cycle control strategy will be considered for various control method.

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

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

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.

Deterioration of mechanical properties of high strength structural steel S460N under transient state fire condition

International Nuclear Information System (INIS)

Qiang, Xuhong; Bijlaard, Frans S.K.; Kolstein, Henk

2012-01-01

Highlights: ► Mechanical properties of S460N under transient state fire condition are obtained. ► Elevated-temperature mechanical properties of steels are dependent on steel grades. ► No design standard is applicable to HSS S460N under transient state fire condition. ► Specific statements on various HSS in fire should be proposed in design standards. ► Research results offer accurate material property for structural design engineers. -- Abstract: 911 World Trade Centre Tragedy put fire safety of constructional steel structures into question. Since then, more and more research attention has been paid to the elevated-temperature mechanical properties of structural steels, which is a critical basis of evaluating the fire performance of steel structures. In the literature the available mechanical properties of structural steels under fire conditions were mainly obtained from steady state test method, as steady state test method is easier to perform than transient state test method and offers stress–strain curves directly. However, the transient state fire condition is considered to be more realistic to represent the real condition when constructions are exposed to fire. In order to reveal the deterioration of mechanical properties of the commonly used high strength structural steel S460N under transient state fire condition, tensile tests were conducted under various constant stress levels up to 800 MPa. The reduction factors of elastic modulus, yield and ultimate strengths of S460N under transient state fire condition were obtained and compared with current leading design standards and available literature. The application of such accurate elevated-temperature mechanical properties reduction factors of S460N can ensure a safe fire-resistance design and evaluation of steel structures with high strength steel S460N under transient state fire condition. This experimental study also supports other relative research on fire performance of steel structures with

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)

Transient Dynamic Mechanical Analysis of Resilin-based Elastomeric Hydrogels

Science.gov (United States)

Li, Linqing; Kiick, Kristi

2014-04-01

The outstanding high-frequency properties of emerging resilin-like polypeptides (RLPs) have motivated their development for vocal fold tissue regeneration and other applications. Recombinant RLP hydrogels show efficient gelation, tunable mechanical properties, and display excellent extensibility, but little has been reported about their transient mechanical properties. In this manuscript, we describe the transient mechanical behavior of new RLP hydrogels investigated via both sinusoidal oscillatory shear deformation and uniaxial tensile testing. Oscillatory stress relaxation and creep experiments confirm that RLP-based hydrogels display significantly reduced stress relaxation and improved strain recovery compared to PEG-based control hydrogels. Uniaxial tensile testing confirms the negligible hysteresis, reversible elasticity and superior resilience (up to 98%) of hydrated RLP hydrogels, with Young’s modulus values that compare favorably with those previously reported for resilin and that mimic the tensile properties of the vocal fold ligament at low strain (engineering applications, of a range of RLP hydrogels.

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

Thermal analysis of LOFT modular DTT for LOCE transient

International Nuclear Information System (INIS)

Martin, C.M.

1978-01-01

A thermal analysis was performed on the LOFT modular drag-disc turbine transducer (MDTT) modular assembly. The purpose of this analysis was to determine the maximum temperature difference between the MDTT shroud and end cap during a LOCE. This temperature difference is needed for stress analysis of the MDTT endcap to fairing welds. The thermal analysis was done using TRIPLE, a three dimensional finite element code. A three dimensional model of the MDTT was made and transient temperature solutions were found for the different MDTT locations. The fluid temperature transients used for the solutions at all locations were from RELAP4 predictions of the LOFT L2-4 test which is considered the most severe temperature transient. Results of these calculations show the maximum temperature difference is 92 0 C (165 0 F) and occurs in the intact loop cold leg. This value and those found at other locations, are evaluated from the best available RELAP predicted temperatures during a nuclear LOCE

Nuclear physics at PEP: First test and future plans

International Nuclear Information System (INIS)

Van Bibber, K.; Dietrich, F.S.; Melnikoff, S.O.

1986-09-01

A test run of internal target nuclear physics at the PEP storage ring is described. The Time Projection Chamber (TPC-2γ detector) was used to detect the inelastically scattered electron and complete hadronic final state in the interaction of 14.5 GeV electrons with D 2 , Ar and Xe gas targets. The data comprise mostly low-x low-Q 2 events, but some deep inelastic scattering as well. The future possibilities of a dedicated nuclear physics program at PEP are outlined. 15 refs., 25 figs

RAPIDLY RISING TRANSIENTS IN THE SUPERNOVA—SUPERLUMINOUS SUPERNOVA GAP

Energy Technology Data Exchange (ETDEWEB)

Arcavi, Iair; Howell, D. Andrew [Las Cumbres Observatory Global Telescope, 6740 Cortona Dr., Suite 102, Goleta, CA 93111 (United States); Wolf, William M. [Department of Physics, University of California, Santa Barbara, CA 93106 (United States); Bildsten, Lars; McCully, Curtis; Valenti, Stefano [Kavli Institute for Theoretical Physics, University of California, Santa Barbara, CA 93106 (United States); Leloudas, Giorgos; Gal-Yam, Avishay; Katz, Boaz [Department of Particle Physics and Astrophysics, The Weizmann Institute of Science, Rehovot, 76100 (Israel); Hardin, Delphine; Astier, Pierre; Balland, Cristophe [LPNHE, CNRS-IN2P3 and University of Paris VI and VII, F-75005 Paris (France); Prajs, Szymon; Sullivan, Mark [School of Physics and Astronomy, University of Southampton, Southampton, SO17 1BJ (United Kingdom); Perley, Daniel A. [Dark Cosmology Centre, Niels Bohr Institute, University of Copenhagen, Juliane Maries Vej 30, DK-2100 Copenhagen (Denmark); Svirski, Gilad [Racah Institute for Physics, The Hebrew University, Jerusalem 91904 (Israel); Cenko, S. Bradley [Astrophysics Science Division, NASA Goddard Space Flight Center, Mail Code 661, Greenbelt, MD 20771 (United States); Lidman, Chris [Australian Astronomical Observatory, P.O. Box 915, North Ryde, NSW 1670 (Australia); Carlberg, Ray G. [Department of Astronomy and Astrophysics, University of Toronto, 50 St. George Street, Toronto, ON M5S 3H8 (Canada); Conley, Alex, E-mail: iarcavi@lcogt.net [Center for Astrophysics and Space Astronomy, University of Colorado, 389 UCB, Boulder, CO 80309-389 (United States); and others

2016-03-01

We present observations of four rapidly rising (t{sub rise} ≈ 10 days) transients with peak luminosities between those of supernovae (SNe) and superluminous SNe (M{sub peak} ≈ −20)—one discovered and followed by the Palomar Transient Factory (PTF) and three by the Supernova Legacy Survey. The light curves resemble those of SN 2011kl, recently shown to be associated with an ultra-long-duration gamma-ray burst (GRB), though no GRB was seen to accompany our SNe. The rapid rise to a luminous peak places these events in a unique part of SN phase space, challenging standard SN emission mechanisms. Spectra of the PTF event formally classify it as an SN II due to broad Hα emission, but an unusual absorption feature, which can be interpreted as either high velocity Hα (though deeper than in previously known cases) or Si ii (as seen in SNe Ia), is also observed. We find that existing models of white dwarf detonations, CSM interaction, shock breakout in a wind (or steeper CSM), and magnetar spin down cannot readily explain the observations. We consider the possibility that a “Type 1.5 SN” scenario could be the origin of our events. More detailed models for these kinds of transients and more constraining observations of future such events should help to better determine their nature.

Analytic models for fuel pin transient performance

International Nuclear Information System (INIS)

Bard, F.E.; Fox, G.L.; Washburn, D.F.; Hanson, J.E.

1976-09-01

HEDL's ability to analyze various mechanisms that operate within a fuel pin has progressed substantially through development of codes such as PECTCLAD, which solves cladding response, and DSTRESS, which solves fuel response. The PECTCLAD results show good correlation with a variety of mechanical tests on cladding material and also demonstrate the significance of cladding strength when applying the life fraction rule. The DSTRESS results have shown that fuel deforms sufficiently during overpower transient tests that available volumes are filled, whether in the form of a central cavity or start-up cracks

MHD aspects of coronal transients

International Nuclear Information System (INIS)

Anzer, U.

1979-10-01

If one defines coronal transients as events which occur in the solar corona on rapid time scales (< approx. several hours) then one would have to include a large variety of solar phenomena: flares, sprays, erupting prominences, X-ray transients, white light transients, etc. Here we shall focus our attention on the latter two phenomena. (orig.) 891 WL/orig. 892 RDG

Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

International Nuclear Information System (INIS)

Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

2012-01-01

FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

Climate variability and vadose zone controls on damping of transient recharge

Science.gov (United States)

Corona, Claudia R.; Gurdak, Jason J.; Dickinson, Jesse; Ferré, T.P.A.; Maurer, Edwin P.

2017-01-01

Increasing demand on groundwater resources motivates understanding of the controls on recharge dynamics so model predictions under current and future climate may improve. Here we address questions about the nonlinear behavior of flux variability in the vadose zone that may explain previously reported teleconnections between global-scale climate variability and fluctuations in groundwater levels. We use hundreds of HYDRUS-1D simulations in a sensitivity analysis approach to evaluate the damping depth of transient recharge over a range of periodic boundary conditions and vadose zone geometries and hydraulic parameters that are representative of aquifer systems of the conterminous United States (U.S). Although the models were parameterized based on U.S. aquifers, findings from this study are applicable elsewhere that have mean recharge rates between 3.65 and 730 mm yr–1. We find that mean infiltration flux, period of time varying infiltration, and hydraulic conductivity are statistically significant predictors of damping depth. The resulting framework explains why some periodic infiltration fluxes associated with climate variability dampen with depth in the vadose zone, resulting in steady-state recharge, while other periodic surface fluxes do not dampen with depth, resulting in transient recharge. We find that transient recharge in response to the climate variability patterns could be detected at the depths of water levels in most U.S. aquifers. Our findings indicate that the damping behavior of transient infiltration fluxes is linear across soil layers for a range of texture combinations. The implications are that relatively simple, homogeneous models of the vadose zone may provide reasonable estimates of the damping depth of climate-varying transient recharge in some complex, layered vadose zone profiles.

Prognostic significance of transient myocardial ischaemia after first acute myocardial infarction: five year follow up study

DEFF Research Database (Denmark)

Mickley, H; Nielsen, J R; Berning, J

1995-01-01

infarction. MAIN OUTCOME MEASURES: Relation of ambulatory ST segment depression, exercise test variables, and left ventricular ejection fraction to subsequent objective (cardiac death or myocardial infarction) or subjective (need for coronary revascularisation) events. RESULTS: 23 of the 123 patients had...... an association between transient ST segment depression and an adverse long term outcome was found (Kaplan-Meier analysis; P = 0.004). The presence of exercise induced angina identified a similar proportion of patients with a poor prognosis (Kaplan-Meier analysis; P ... ST segment depression had high specificity but poor sensitivity. The presence of exercise induced ST segment depression was of no value in predicting combined cardiac events. Indeed, patients without exertional ST segment depression were at increased risk of future objective end points (Kaplan...

What should we want to know about our future? A Kantian view on predictive genetic testing.

Science.gov (United States)

Heinrichs, Bert

2005-01-01

Recent advances in genomic research have led to the development of new diagnostic tools, including tests which make it possible to predict the future occurrence of monogenetic diseases (e.g. Chorea Huntington) or to determine increased susceptibilities to the future development of more complex diseases (e.g. breast cancer). The use of such tests raises a number of ethical, legal and social issues which are usually discussed in terms of rights. However, in the context of predictive genetic tests a key question arises which lies beyond the concept of rights, namely, What should we want to know about our future? In the following I shall discuss this question against the background of Kant's Doctrine of Virtue. It will be demonstrated that the system of duties of virtue that Kant elaborates in the second part of his Metaphysics of Morals offers a theoretical framework for addressing the question of a proper scope of future knowledge as provided by genetic tests. This approach can serve as a source of moral guidance complementary to a justice perspective. It does, however, not rest on the-rather problematic--claim to be able to define what the "good life" is.

BWR Full Integral Simulation Test (FIST). Phase I test results

International Nuclear Information System (INIS)

Hwang, W.S.; Alamgir, M.; Sutherland, W.A.

1984-09-01

A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report

Transient-Switch-Signal Suppressor

Science.gov (United States)

Bozeman, Richard J., Jr.

1995-01-01

Circuit delays transmission of switch-opening or switch-closing signal until after preset suppression time. Used to prevent transmission of undesired momentary switch signal. Basic mode of operation simple. Beginning of switch signal initiates timing sequence. If switch signal persists after preset suppression time, circuit transmits switch signal to external circuitry. If switch signal no longer present after suppression time, switch signal deemed transient, and circuit does not pass signal on to external circuitry, as though no transient switch signal. Suppression time preset at value large enough to allow for damping of underlying pressure wave or other mechanical transient.

Summary of transient analysis

International Nuclear Information System (INIS)

Saha, P.

1984-01-01

This chapter reviews the papers on the pressurized water reactor (PWR) and boiling water reactor (BWR) transient analyses given at the American Nuclear Society Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors. Most of the papers were based on the systems calculations performed using the TRAC-PWR, RELAP5 and RETRAN codes. The status of the nuclear industry in the code applications area is discussed. It is concluded that even though comprehensive computer codes are available for plant transient analysis, there is still a need to exercise engineering judgment, simpler tools and even hand calculations to supplement these codes

Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

Energy Technology Data Exchange (ETDEWEB)

Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

2005-09-27

The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

Thermal–stress analysis on the crack formation of tungsten during fusion relevant transient heat loads

Directory of Open Access Journals (Sweden)

Changjun Li

2017-12-01

Full Text Available In the future fusion devices, ELMs-induced transient heat flux may lead to the surface cracking of tungsten (W based plasma-facing materials (PFMs. In theory, the cracking is related to the material fracture toughness and the thermal stress-strain caused by transient heat flux. In this paper, a finite element model was successfully built to realize a theoretical semi infinite space. The temperature and stress-strain distribution as well as evolution of W during a single heating-cooling cycle of transient heat flux were simulated and analyzed. It showed that the generation of plastic deformation during the brittle temperature range between room temperature and DBTT (ductile to brittle transition temperature, ∼400 °C caused the cracking of W during the cooling phase. The cracking threshold for W under transient heat flux was successfully obtained by finite element analysis, to some extent, in consistent with the similar experimental results. Both the heat flux factors (FHF = P·t0.5 and the maximum surface temperatures at cracking thresholds were almost invariant for the transient heat fluxes with different pulse widths and temporal distributions. This method not only identified the theoretical conclusion but also obtained the detail values for W with actual temperature-dependent properties.

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

Objective Evaluation of the Audibility of Transient Errors in an Adaptive A/D Conversion Channel

DEFF Research Database (Denmark)

Marker-Villumsen, Niels; Jørgensen, Ivan Harald Holger; Bruun, Erik

2014-01-01

An adaptive analog-to-digital conversion channel for audio, using automatic gain control, generates transient errors that may be audible. Evaluating the audibility of such errors requires subjective evaluation using listening tests. From an electrical circuit design point-of-view this is not feas......An adaptive analog-to-digital conversion channel for audio, using automatic gain control, generates transient errors that may be audible. Evaluating the audibility of such errors requires subjective evaluation using listening tests. From an electrical circuit design point...

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).

Present status and future perspective of research and test reactors in JAERI

International Nuclear Information System (INIS)

Baba, Osamu; Kaieda, Keisuke

1999-01-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

Present status and future perspective of research and test reactors in JAERI

Energy Technology Data Exchange (ETDEWEB)

Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

1999-08-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

500 MHz transient digitizers based on GaAs CCDs

International Nuclear Information System (INIS)

Bryman, D.; Cresswell, J.V.; LeNoble, M.; Poutissou, R.

1990-10-01

A wide bandwidth transient digitizer based on a recently produced gallium arsenide charged coupled device is under development. The CCDs have 128 pixels and operate at 500 MHz. Initial testing of prototype modules in Experiment 787 at Brookhaven National Laboratory is reported. (Author) (8 refs., 10 figs.)

Stirling Convertor Performance Mapping Test Results for Future Radioisotope Power Systems

Science.gov (United States)

Qiu, Songgang; Peterson, Allen A.; Faultersack, Franklyn D.; Redinger, Darin L.; Augenblick, John E.

2004-02-01

Long-life radioisotope-fueled generators based on free-piston Stirling convertors are an energy-conversion solution for future space applications. The high efficiency of Stirling machines makes them more attractive than the thermoelectric generators currently used in space. Stirling Technology Company (STC) has been performance-testing its Stirling generators to provide data for potential system integration contractors. This paper describes the most recent test results from the STC RemoteGen™ 55 W-class Stirling generators (RG-55). Comparisons are made between the new data and previous Stirling thermodynamic simulation models. Performance-mapping tests are presented including variations in: internal charge pressure, cold end temperature, hot end temperature, alternator temperature, input power, and variation of control voltage.

Transient turbulent heat transfer for heating of water in a short vertical tube

International Nuclear Information System (INIS)

Hata, Koichi; Kai, Naoto; Shirai, Yasuyuki; Masuzaki, Suguru

2011-01-01

The transient turbulent heat transfer coefficients in a short vertical Platinum test tube were systematically measured for the flow velocities (u=4.0 to 13.6 m/s), the inlet liquid temperatures (T in =296.93 to 304.81 K), the inlet pressures (P in =794.39 to 858.27 kPa) and the increasing heat inputs (Q 0 exp(t/τ), exponential periods, τ, of 18.6 ms to 25.7 s) by an experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The Platinum test tubes of test tube inner diameters (d=3 and 6 mm), heated lengths (L=66.5 and 69.6 mm), effective lengths (L eff =56.7 and 59.2 mm), ratios of heated length to inner diameter (L/d=22.16 and 11.6), ratios of effective length to inner diameter (L eff /d=18.9 and 9.87) and wall thickness (δ=0.5 and 0.4 mm) with average surface roughness (Ra=0.40 and 0.45 μm) were used in this work. The surface heat fluxes between the two potential taps were given the difference between the heat generation rate per unit surface area and the rate of change of energy storage in the test tube obtained from the faired average temperature versus time curve. The heater inner surface temperature between the two potential taps was also obtained by solving the unsteady heat conduction equation in the test tube under the conditions of measured average temperature and heat generation rate per unit surface area of the test tube. The transient turbulent heat transfer data for Platinum test tubes were compared with the values calculated by authors' correlation for the steady state turbulent heat transfer. The influence of inner diameter (d), ratio of effective length to inner diameter (L eff /d), flow velocity (u) and exponential period (τ) on the transient turbulent heat transfer is investigated into details and the widely and precisely predictable correlation of the transient turbulent heat transfer for heating of water in a short vertical tube is given based on the experimental data and authors' studies for the

Transient turbulent heat transfer for heating of water in a short vertical tube

International Nuclear Information System (INIS)

Hata, Koichi; Kai, Naoto; Shirai, Yasuyuki; Masuzaki, Suguru

2011-01-01

The transient turbulent heat transfer coefficients in a short vertical Platinum test tube were systematically measured for the flow velocities (u=4.0 to 13.6 m/s), the inlet liquid temperatures (T in =296.93 to 304.81 K), the inlet pressures (P in =794.39 to 858.27 kPa) and the increasing heat inputs (Q 0 exp(t/τ), exponential periods, τ, of 18.6 ms to 25.7 s) by an experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The Platinum test tubes of test tube inner diameters (d=3 and 6 mm), heated lengths (L=66.5 and 69.6 mm), effective lengths (L eff =56.7 and 59.2 mm), ratios of heated length to inner diameter (L/d=22.16 and 11.6), ratios of effective length to inner diameter (L eff /d=18.9 and 9.87) and wall thickness (δ=0.5 and 0.4 mm) with average surface roughness (Ra=0.40 and 0.45 μm) were used in this work. The surface heat fluxes between the two potential taps were given the difference between the heat generation rate per unit surface area and the rate of change of energy storage in the test tube obtained from the faired average temperature versus time curve. The heater inner surface temperature between the two potential taps was also obtained by solving the unsteady heat conduction equation in the test tube under the conditions of measured average temperature and heat generation rate per unit surface area of the test tube. The transient turbulent heat transfer data for Platinum test tubes were compared with the values calculated by authors' correlation for the steady state turbulent heat transfer. The influence of inner diameter (d), ratio of effective length to inner diameter (L eff /d), flow velocity (u) and exponential period (τ) on the transient turbulent heat transfer is investigated into details and the widely and precisely predictable correlation of the transient turbulent heat transfer for heating of water in a short vertical tube is given based on the experimental data and authors' studies for the

Irradiation facilitates at the advanced test reactor

International Nuclear Information System (INIS)

Grover, Blaine S.

2006-01-01

The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

PWR systems transient analysis

International Nuclear Information System (INIS)

Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

1985-01-01

Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

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

Simulation of Thermal Transients using CSMP

International Nuclear Information System (INIS)

Konuk, A.A.

1981-01-01

A mathematical model has been developed to simulate thermal transientes for the Hellum Loop of the 'Instituto de Pesquisas Energeticas e Nuleares', Sao Paulo. The model is based on the energy equation applied to the various components of the loop. The non-linear system of first order ordinary differential equation and algebraic equations has been solved using IBM'S 'System/360-Continuous System Modeling Program-CSMP'. The model has been tested satisfactory with experimental results. (Author) [pt

Effect of transient liquid phase (TLP) bonding on the ductility of a Ni-base single crystal superalloy in a stress rupture test

International Nuclear Information System (INIS)

Liu, J.D.; Jin, T.; Zhao, N.R.; Wang, Z.H.; Sun, X.F.; Guan, H.R.; Hu, Z.Q.

2008-01-01

A Ni-base single crystal superalloy was transient liquid phase (TLP) bonded using a Ni-Cr-B amorphous foil at 1230 deg. C for 8 h. Stress rupture tests of the TLP joint and a matrix sample were carried out at 982 deg. C/248 MPa and 1010 deg. C/248 MPa. The microstructures and fracture surfaces were studied using scanning electron microscopy (SEM). Transmission electron microscopy (TEM) investigations were performed after creep rupture testing to examine the deformation substructures. The results show that the stress rupture ductility of TLP joints is significantly decreased compared to the matrix sample. This reduction of the ductility of TLP joints can be attributed to solid solution strengthening by boron atoms, subgrain boundaries formed in the bonding zone and the concentration of creep cavities formed during the last stage of the stress rupture test

Transient osteoporosis of the hip

International Nuclear Information System (INIS)

McWalter, Patricia; Hassan Ahmed

2007-01-01

Transient osteoporosis of the hip is an uncommon cause of hip pain, mostly affecting healthy middle-aged men and also women in the third trimester of pregnancy. We present a case of transient osteoporosis of the hip in a 33-year-old non-pregnant female patient. This case highlights the importance of considering a diagnosis of transient osteoporosis of the hip in patients who present with hip pain. (author)

Effect of a Chloride Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 2)

International Nuclear Information System (INIS)

Ritter, S.; Seifert, H.-P.

2002-11-01

Within the CASTOC-project (5 t h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the second test of working package (WP) 3 with a NaCl transient, performed at Paul Scherrer Institut (PSI). In the first part of the experiment, an actively growing EAC crack with a crack growth rate (CGR) in the range of the 'low-sulphur SCC line' of the GE-model was generated by periodical partial unloading (PPU) in oxygenated high-temperature, high-purity water (T = 288 o C, DO = 8 ppm). Then a chloride transient of 49 ppb Cl - was applied for ∼40 h. After this transient, the load level was reduced and the loading conditions were changed to pure cyclic loading. Thereupon a second transient with a chloride concentration of 49 ppb was applied. In both RPV steels, the first chloride transient of 49 ppb Cl - resulted in an acceleration of the EAC crack growth by more than one order of magnitude and in fast, stationary SCC crack growth during the constant load phase of the PPU cycles at K I values 1/2 . 3 h after adding chloride to the high-purity water, the EAC CGR started to increase in the high-sulphur RPV steel during the constant load phase of a PPU cycle and after 20 h a stationary EAC CGR value in the range of the 'high-sulphur SCC curve' of the GE-model was reached. After 5 h in high-purity water, the crack growth began to slow down after a partial unloading cycle and 15 h later it reached again a stationary CGR value in the range of the 'low-sulphur SCC curve' of the GE-model. The second chloride transient did not result in an acceleration of the crack growth in both investigated specimens. This was explained by crack closure effects, which occurred in both specimens after the reduction of the load. The CGR

Suppression of thermal transients in advanced LIGO interferometers using CO2 laser preheating

Science.gov (United States)

Jaberian Hamedan, V.; Zhao, C.; Ju, L.; Blair, C.; Blair, D. G.

2018-06-01

In high optical power interferometric gravitational wave detectors, such as Advanced LIGO, the thermal effects due to optical absorption in the mirror coatings and the slow thermal response of fused silica substrate cause time dependent changes in the mirror profile. After locking, high optical power builds up in the arm cavities. Absorption induced heating causes optical cavity transverse mode frequencies to drift over a period of hours, relative to the fundamental mode. At high optical power this can cause time dependent transient parametric instability, which can lead to interferometer disfunction. In this paper, we model the use of CO2 laser heating designed to enable the interferometer to be maintained in a thermal condition such that transient changes in the mirrors are greatly reduced. This can minimize transient parametric instability and compensate dark port power fluctuations. Modeling results are presented for both single compensation where a CO2 laser acting on one test mass per cavity, and double compensation using one CO2 laser for each test mass. Using parameters of the LIGO Hanford Observatory X-arm as an example, single compensation allows the maximum mode frequency shift to be limited to 6% of its uncompensated value. However, single compensation causes transient degradation of the contrast defect. Double compensation minimise contrast defect degradation and reduces transients to less than 1% if the CO2 laser spot is positioned within 2 mm of the cavity beam position.

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)

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

RETRAN operational transient analysis of the Big Rock Point plant boiling water reactor

International Nuclear Information System (INIS)

Sawtelle, G.R.; Atchison, J.D.; Farman, R.F.; VandeWalle, D.J.; Bazydlo, H.G.

1983-01-01

Energy Incorporated used the RETRAN computer code to model and calculate nine Consumers Power Company Big Rock Point Nuclear Power Plant transients. RETRAN, a best-estimate, one-dimensional, homogeneous-flow thermal-equilibrium code, is applicable to FSAR Chapter 15 transients for Conditions 1 through IV. The BWR analyses were performed in accordance with USNRC Standard Review Plan criteria and in response to the USNRC Systematic Evaluation Program. The RETRAN Big Rock Point model was verified by comparison to plant startup test data. This paper discusses the unique modeling techniques used in RETRAN to model this steam-drum-type BWR. Transient analyses results are also presented

Tests of Scintillator+WLS Strips for Muon System at Future Colliders

Energy Technology Data Exchange (ETDEWEB)

Denisov, Dmitri [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Evdokimov, Valery [Inst. for High Energy Physics (IHEP), Protvino (Russian Federation); Lukić, Strahinja [Univ. of Belgrade (Serbia)

2015-10-11

Prototype scintilator+WLS strips with SiPM readout for muon system at future colliders were tested for light yield, time resolution and position resolution. Depending on the configuration, light yield of up to 36 photoelectrons per muon per SiPM has been achieved, as well as time resolution of 0.5 ns and position resolution of ~ 7 cm.

Quantitative assessment of the BETHSY 6.9c test simulation

International Nuclear Information System (INIS)

Hrvatin, S.; Prosek, A.

2000-01-01

In the field of nuclear engineering, complex thermal-hydraulic computer codes are used to simulate and predict various transients in nuclear power plants. These computer codes are validated for overall system simulation by using experimental results, obtained on the integral test facilities. A post-test calculation of BETHSY 6.9c test with the RELAP5/MOD3.2 computer code has been performed in order to improve the input model in the future. The qualitative comparison of the results showed that most of the relevant parameters are predicted reasonably well. The quantitative assessment of the results was performed using the so-called Fast Fourier Transform Based Methodology. The FFTBM delineates and quantifies differences between calculated and experimental parameters in the frequency domain. The analysis showed that the the code calculations yield acceptable results. However, the primary pressure acceptability criterion is not fulfilled. This indicates that primary pressure calculation at low pressures is less accurate than at typical transient conditions. In general, it can be concluded that the RELAP5/MOD3.2 computer code can be used to analyze midloop operation at low power and pressure conditions. (author)

A COMETHE version with transient capability

International Nuclear Information System (INIS)

Vliet, J. van; Lebon, G.; Mathieu, P.

1980-01-01

A version of the COMETHE code is under development to simulate transient situations. This paper focuses on some aspects of the transient heat transfer models. Initially the coupling between transient heat transfer and other thermomechanical models is discussed. An estimation of the thermal characteristic times shows that the cladding temperatures are often in quasi-steady state. In order to reduce the computing time, calculations are therefore switched from a transient to a quasi-static numerical procedure as soon as such a quasi-equilibrium is detected. The temperature calculation is performed by use of the Lebon-Lambermont restricted variational principle, with piecewise polynoms as trial functions. The method has been checked by comparison with some exact results and yields good agreement for transient as well as for quasi-static situations. This method therefore provides a valuable tool for the simulation of the transient behaviour of nuclear reactor fuel rods. (orig.)

System for measuring of air concentration in air-steam mixture during the transients

International Nuclear Information System (INIS)

Gorbenko, Gennady A.; Gakal, Pavlo G.; Epifanov, Konstantin S.; Osokin, Gennady V.; Smirnov, Sergey V.

2006-01-01

Description of system for air concentration measuring in air-steam mixture during the transients is represented. Air concentration measuring is based on discrete sampling method. The measuring system consists of sampler, transport pipeline, distributor and six measuring vessels. From the sampler air-steam mixture comes to distributor through transport pipeline and fills consecutively the measuring vessels. The true air concentration in place of measurement was defined based on measured air concentration in samples taken from measuring vessels. For this purpose, the mathematical model of transients in measuring system was developed. Air concentration transient in air-steam mixture in place of measurement was described in mathematical model by air concentration time-dependent function. The function parameters were defined based on air concentration measured in samples taken from measuring vessels. Estimated error of air concentration identification was about 10%. Measuring system was used in experiments on EREC BKV-213 test facility intended for testing of VVER-440/V-213 reactor barbotage-vacuum system

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

Research on the improvement of nuclear safety -Development of a nuclear power plant system analysis code TASS (Transient and setpoint simulation)

Energy Technology Data Exchange (ETDEWEB)

Shim, Suk Koo; Jang, Won Pyo; Kim, Heui Chul; Kim, Kyung Doo; Lee, Sung Jae; Hah, Kyooi Suk; Song, Soon Jah; Um, Kil Sub; Yoon, Han Yung; Kim, Doo Il; Yoo, Hyung Keun; Choi, Jae Don; Lee, Byung Il; Kim, Jung Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-07-01

During the third year of the project the development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the Westinghouse and CE type operating reactors as well as the PWR reactors under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the YGN-3/4 FSAR transients, Kori-3 loss of AC power transient, plant data, Kori-4 load rejection and YGN-3 startup test data as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best estimate RELAP5/MOD 3.1 calculation for the YGN-3/4 FASR transients and shows its capability in simulating plant transient and startup data as well as the thermal hydraulic transient test data. Topical reports on TASS 1.0 code have been prepared and will be submitted to Korea Institute of Nuclear Safety for its licensing application to Westinghouse and CE type PWR transient analyses. The development of TASS 2.0 code has been head started in this year to timely utilize the TASS 2.0 code for the KNGR design certification. 65 figs, 30 tabs, 44 refs. (Author).

Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

Energy Technology Data Exchange (ETDEWEB)

Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

2010-10-15

This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

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

Thermohydraulics in rod bundles and critical heat flux in transient conditions in a tube

International Nuclear Information System (INIS)

Courtaud, M.; Roumy, R.

1975-01-01

After the determination of the scaling factor of Stevens's similitude for the pressure range of pressurized water vectors by comparison of critical heat flux data in from and in water, some examples of studies performed with freon are shown. The efficiency of the mixing vanes of spacer grids has been determined on the mixing phenomenon in single phase on critical heat flux. A calculation performed with the code FLICA using subchannel analysis on freon data transposed in water is in good agreement with the experiment. The influence of the number of spacer grids has been also shown. Critical heat fluxes have been determined in water at 140 bar in steady state and transient conditions on two tubular test sections. During the transient tests the flow rate was reduced by half in 0.5 seconds and the reincreased heat flux and inlet temperature remaining constant. These tests have shown the validity of the method which consists in using a critical heat flux correlation determined in steady state conditions applied with local transient conditions of enthalpy and mass velocity computed with the FLICA code [fr

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

International Nuclear Information System (INIS)

Daw, J.E.; Knudson, D.L.; Villard, J.F.; Liothin, J.; Destouches, C.; Rempe, J.L.; Matheron, P.; Lambert, T.

2015-01-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

Energy Technology Data Exchange (ETDEWEB)

Daw, J.E.; Knudson, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415, (United States); Villard, J.F.; Liothin, J.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Rempe, J.L. [Rempe and Associates, LLC, Idaho Falls, ID, 83404 (United States); Matheron, P. [CEA, DEN, DEC, Uranium Fuels Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Lambert, T. [CEA, DEN, DEC, Innovative Fuel Design and Irradiation Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France)

2015-07-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

CEDNBR: a computer code for transient thermal margin analysis of a reactor core

International Nuclear Information System (INIS)

Shesler, A.T.; Lehmann, C.R.

1976-09-01

The report describes the CEDNBR computer code. This code was developed for the transient thermal analysis of a pressurized water reactor core or a critical heat flux test. Included are the code structure, conservation equations, and correlations utilized by CEDNBR. The methods of modelling a reactor core and hot channel and a CHF test are presented. Comparisons of CEDNBR calculations are made with both empirical pressure loss data and simulated loss of flow test data. The code solves the one-dimensional conservation of mass, energy, and momentum equations and the equation of state for the fluid for either steady-state or transient conditions. Tabular time dependent functions of inlet temperatures, pressure, mass velocity, axial heat flux distributions, normalized heat flux, radial peaking factors, and incremental mixing factors are required input to the code. Transient effects are included in the calculation of enthalpy rise and fluid properties. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by applying a Critical Heat Flux (CHF) correlation to the computed local fluid properties. A code user's guide is provided for preparing input to the code. In addition, descriptions of the sub-routines used by CEDNBR are given

Study of transient burnout under flow reduction condition

International Nuclear Information System (INIS)

Iwamura, Takamichi

1986-09-01

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

Fission product transport and behavior during two postulated loss of flow transients in the air

International Nuclear Information System (INIS)

Adams, J.P.; Carboneau, M.L.

1991-01-01

This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10 -5 and 10 -7 per reactor year for LCP15 and LPP9, respectively

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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.

Research of the transient management in TQNPC

International Nuclear Information System (INIS)

Guo Longzhang; Lin Chuanqing

2008-01-01

Transient management is the basic technical subject in nuclear power plant. Since the Third Qinshan nuclear power company (TQNPC) successful completes the commissioning in 2003, the transient management work start at the transient management item selection and the flow definition. Now TQNPC have a complete transient management system and the management flow. In the last two years, TNQPC have finished the historic transient data collection for two units, and confirmed that the plant's key systems and equipments are at safe state. The development of the transient management subject would build a reliable foundation for the plant safe operation, plant lifetime management and periodic safety review. (author)

Compressive Transient Imaging

KAUST Repository

Sun, Qilin

2017-04-01

High resolution transient/3D imaging technology is of high interest in both scientific research and commercial application. Nowadays, all of the transient imaging methods suffer from low resolution or time consuming mechanical scanning. We proposed a new method based on TCSPC and Compressive Sensing to achieve a high resolution transient imaging with a several seconds capturing process. Picosecond laser sends a serious of equal interval pulse while synchronized SPAD camera\\'s detecting gate window has a precise phase delay at each cycle. After capturing enough points, we are able to make up a whole signal. By inserting a DMD device into the system, we are able to modulate all the frames of data using binary random patterns to reconstruct a super resolution transient/3D image later. Because the low fill factor of SPAD sensor will make a compressive sensing scenario ill-conditioned, We designed and fabricated a diffractive microlens array. We proposed a new CS reconstruction algorithm which is able to denoise at the same time for the measurements suffering from Poisson noise. Instead of a single SPAD senor, we chose a SPAD array because it can drastically reduce the requirement for the number of measurements and its reconstruction time. Further more, it not easy to reconstruct a high resolution image with only one single sensor while for an array, it just needs to reconstruct small patches and a few measurements. In this thesis, we evaluated the reconstruction methods using both clean measurements and the version corrupted by Poisson noise. The results show how the integration over the layers influence the image quality and our algorithm works well while the measurements suffer from non-trival Poisson noise. It\\'s a breakthrough in the areas of both transient imaging and compressive sensing.

Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

International Nuclear Information System (INIS)

1996-10-01

This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

Work plan: transient release from LMFBR fuel

International Nuclear Information System (INIS)

Kress, T.S.; Parker, G.W.; Fontana, M.H.

1975-09-01

The proposed LMFBR Transient Release Program at ORNL is designed to investigate, by means of ex-reactor experiments and analytical modeling, the release and transport of fuel, fission products, and transuranic elements from fast reactor cores in the event of certain hypothetical accidents. It is desired to experimentally produce energy depositions that are characteristic of severe hypothetical reactor transients by the application of direct electrical current to mixed-oxide fuels under sodium. The experimental program includes tests with and without sodium, investigations of alternative methods of generating fuel and sodium aerosols, the use of UO 2 as a fuel simulant, additions of tracers as fission product simulants, effects of radiation, and under-water and under-sodium efforts to study the behavior of the vapor bubble itself. Analytical modeling will accompany all phases of the program, and the data will be correlated with models developed. 21 references. (auth)

Modeling transient radiation effects in power MOSFETS

International Nuclear Information System (INIS)

Hoffman, J.R.; Hall, W.E.; Dunn, D.E.

1987-01-01

Using standard device specifications and simple assumptions, the transient radiation response of VDMOS MOSFETs can be modeled in a standard circuit analysis program. The device model consists of a body diode, a parasitic bipolar transistor, and elements to simulate high-current reduced breakdown. The attached photocurrent model emulates response to any pulse shape and accounts for bias-dependent depletion regions. The model can be optimized to best fit available test data

A transient absorption study of allophycocyanin

Indian Academy of Sciences (India)

Transient dynamics of allophycocyanin trimers and monomers are observed by using the pump-probe, transient absorption technique. The origin of spectral components of the transient absorption spectra is discussed in terms of both kinetics and spectroscopy. We find that the energy gap between the ground and excited ...

The joy of transient chaos

Energy Technology Data Exchange (ETDEWEB)

Tél, Tamás [Institute for Theoretical Physics, Eötvös University, and MTA-ELTE Theoretical Physics Research Group, Pázmány P. s. 1/A, Budapest H-1117 (Hungary)

2015-09-15

We intend to show that transient chaos is a very appealing, but still not widely appreciated, subfield of nonlinear dynamics. Besides flashing its basic properties and giving a brief overview of the many applications, a few recent transient-chaos-related subjects are introduced in some detail. These include the dynamics of decision making, dispersion, and sedimentation of volcanic ash, doubly transient chaos of undriven autonomous mechanical systems, and a dynamical systems approach to energy absorption or explosion.

The joy of transient chaos.

Science.gov (United States)

Tél, Tamás

2015-09-01

We intend to show that transient chaos is a very appealing, but still not widely appreciated, subfield of nonlinear dynamics. Besides flashing its basic properties and giving a brief overview of the many applications, a few recent transient-chaos-related subjects are introduced in some detail. These include the dynamics of decision making, dispersion, and sedimentation of volcanic ash, doubly transient chaos of undriven autonomous mechanical systems, and a dynamical systems approach to energy absorption or explosion.

The ZTF Bright Transient Survey

Science.gov (United States)

Fremling, C.; Sharma, Y.; Kulkarni, S. R.; Miller, A. A.; Taggart, K.; Perley, D. A.; Gooba, A.

2018-06-01

As a supplement to the Zwicky Transient Facility (ZTF; ATel #11266) public alerts (ATel #11685) we plan to report (following ATel #11615) bright probable supernovae identified in the raw alert stream from the ZTF Northern Sky Survey ("Celestial Cinematography"; see Bellm & Kulkarni, 2017, Nature Astronomy 1, 71) to the Transient Name Server (https://wis-tns.weizmann.ac.il) on a daily basis; the ZTF Bright Transient Survey (BTS; see Kulkarni et al., 2018; arXiv:1710.04223).

Analysis of chemical warfare using a transient semi-Markov formulation.

OpenAIRE

Kierzewski, Michael O.

1988-01-01

Approved for public release; distribution is unlimited This thesis proposes an analytical model to test various assumptions about conventional/chemical warfare. A unit's status in conventional/chemical combat is modeled as states in a semi-Markov chain with transient and absorbing states. The effects of differing chemical threat levels, availability of decontamination assets and assumed personnel degradation rates on expected unit life and capabilities are tested. The ...

Critical power characteristics in 37-rod tight lattice bundles under transient conditions

International Nuclear Information System (INIS)

Liu, Wei; Kureta, Masatoshi; Tamai, Hidesada; Ohnuki, Akira; Akimoto, Hajime

2007-01-01

Critical power characteristics in the postulated abnormal transient processes that may be possibly met in the operation of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) were investigated for the design of the FLWR core. Transient Boiling Transition (BT) tests were carried out using two sets of 37-rod tight lattice rod bundles (rod diameter: 13 mm; rod clearance: 1.3 mm or 1.0 mm) at Japan Atomic Energy Agency (JAEA) under the conditions covering the FLWR operating condition (P ex =7.2 MPa, T in =556 K) for mass velocity G=400-800 kg/(m 2 s). For the postulated power increase and flow decrease transients, no obvious change of the critical power against the steady one was observed. The traditional quasi-steady characteristic was confirmed to be working for the postulated power increase and flow decrease transients. The experiments were analyzed with TRAC-BF1 code, where the JAEA newest critical power correlation for the tight lattice rod bundles was implemented for the BT judgment. The TRAC-BF1 code showed good prediction for the occurrence or the non occurrence of the BT and for the exact BT starting time. The tranditional quasi-steady state prediction of the BT in transient process was confirmed to be applicable for the postulated abnormal transient processes in the tight lattice rod bundles. (author)

Transient simulation of a helical-coil sodium/water steam generator

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Iwashita, T.

1982-01-01

The MINET (Momentum Integral Network) code heat exchanger model was used to analyze transient test data provided by PNC of Japan. Testing of the MINET model is part of a larger effort to facilitate and validate the use of the SSC/MINET code for MONJU plant transient analysis. In MINET, a heat exchanger is modeled using one or more representative tubes, with each tube consisting of the fluid inside the tube, the tube wall, and the fluid outside that is associated with the tube. The heat exchanger tube is divided into one or more axial nodes of equal length. Five time dependent equations are utilized per node, including the wall heat conduction equation and donor-cell differenced conservation of mass and energy equations for the fluids on both sides of the tube. These nodal equations are used to constrain the tube wall temperature, fluid mass flow rates, and fluid enthalpies

Explosive and radio-selected Transients: Transient Astronomy with ...

Indian Academy of Sciences (India)

40

sitive measurements will lead to very accurate mass loss estimation in these supernovae. .... transients are powerful probes of intervening media owing to dispersion ...... A., & Chandra, P. 2011, Nature Communications,. 2, 175. Chakraborti, S.

Test OPTRAN 1-1 results

International Nuclear Information System (INIS)

Martinson, Z.R.

1982-01-01

The objective of the OPT 1-1 Test Series was to evaluate the extent of damage and the threshold for failure during simulated BWR anticipated transients. Four power transient tests with progressively higher power levels were performed with preirradiated fuel rods at power ramp rates as high as 550 kW/m per second. Six separately shrouded fuel rods fabricated by the General Electric Co., and preirradiated in the Monticello BWR to burnups of about 5000 to 23,000 MWd/t were tested, four at a time. Four of the fuel rods were of typical GE 8 x 8 design, except for fuel length (0.75 m). Two of the rods included design modifications to improve their PCI-resistant characteristics. A lengthy fuel conditioning preceded the transient testing of the fuel rods

In Situ Estuarine and Marine Toxicity Testing: A Review, Including Recommendations for Future Use in Ecological Risk Assessment

Science.gov (United States)

2009-09-01

field and microcosms than they do under laboratory test conditions. In the case of tributyltin ( TBT ) exposures in San Diego Bay, he found that...TECHNICAL REPORT 1986 September 2009 In Situ Estuarine and Marine Toxicity Testing A Review, Including Recommendations for Future Use in...Pacific TECHNICAL REPORT 1986 September 2009 In Situ Estuarine and Marine Toxicity Testing A Review, Including Recommendations for Future Use in

The Dynamic Radio Sky: Future Directions at cm/m-Wavelengths

Science.gov (United States)

Bower, Geoffrey C.; Cordes, J.; Croft, S.; Lazio, J.; Lorimer, D.; McLaughlin, M.

2009-01-01

The time domain of the radio wavelength sky has been only sparsely explored. Nevertheless, recent discoveries from limited surveys and serendipitous discoveries indicate that there is much to be found on timescales from nanoseconds to years and at wavelengths from meters to millimeters. These observations have revealed unexpected phenonmena such as rotating radio transients and coherent pulses from brown dwarfs. Additionally, archival studies have revealed an unknown class of radio transients without radio, optical, or high-energy hosts. The current generation of new meter- and centimeter-wave radio telescopes such as the MWA, LWA, PAPER, and ATA will exploit wide fields of view and flexible digital signal processing to systematically explore radio transient parameter space, as well as lay the scientific and technical foundation for the SKA. Known unknowns that will be the target of future transient surveys include orphan gamma-ray burst afterglows, radio supernovae, tidally-disrupted stars, flare stars, and magnetars.

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

Electromagnetic transients in power cables

CERN Document Server

da Silva, Filipe Faria

2013-01-01

From the more basic concepts to the most advanced ones where long and laborious simulation models are required, Electromagnetic Transients in Power Cables provides a thorough insight into the study of electromagnetic transients and underground power cables. Explanations and demonstrations of different electromagnetic transient phenomena are provided, from simple lumped-parameter circuits to complex cable-based high voltage networks, as well as instructions on how to model the cables.Supported throughout by illustrations, circuit diagrams and simulation results, each chapter contains exercises,

Transient multimessenger astronomy with gravitational waves

International Nuclear Information System (INIS)

Marka, S

2011-01-01

Comprehensive multimessenger astronomy with gravitational waves is a pioneering field bringing us interesting results and presenting us with exciting challenges for the future. During the era of the operation of advanced interferometric gravitational wave detectors, we will have the opportunity to investigate sources of gravitational waves that are also expected to be observable through other messengers, such as gamma rays, x-rays, optical, radio, and/or neutrino emission. Multimessenger searches for gravitational waves with the LIGO-GEO600-Virgo interferometer network have already produced insights on cosmic events and it is expected that the simultaneous observation of electromagnetic or neutrino emission could be a crucial aspect for the first direct detection of gravitational waves in the future. Trigger time, direction and expected frequency range enhances our ability to search for gravitational wave signatures with amplitudes closer to the noise floor of the detector. Furthermore, multimessenger observations will enable the extraction of otherwise unaccessible scientific insight. We summarize the status of transient multimessenger detection efforts as well as mention some of the open questions that might be resolved by advanced or third generation gravitational wave detector networks.

Recrystallization and grain growth induced by ELMs-like transient heat loads in deformed tungsten samples

Science.gov (United States)

Suslova, A.; El-Atwani, O.; Sagapuram, D.; Harilal, S. S.; Hassanein, A.

2014-11-01

Tungsten has been chosen as the main candidate for plasma facing components (PFCs) due to its superior properties under extreme operating conditions in future nuclear fusion reactors such as ITER. One of the serious issues for PFCs is the high heat load during transient events such as ELMs and disruption in the reactor. Recrystallization and grain size growth in PFC materials caused by transients are undesirable changes in the material, since the isotropic microstructure developed after recrystallization exhibits a higher ductile-to-brittle transition temperature which increases with the grain size, a lower thermal shock fatigue resistance, a lower mechanical strength, and an increased surface roughening. The current work was focused on careful determination of the threshold parameters for surface recrystallization, grain growth rate, and thermal shock fatigue resistance under ELM-like transient heat events. Transient heat loads were simulated using long pulse laser beams for two different grades of ultrafine-grained tungsten. It was observed that cold rolled tungsten demonstrated better power handling capabilities and higher thermal stress fatigue resistance compared to severely deformed tungsten. Higher recrystallization threshold, slower grain growth, and lower degree of surface roughening were observed in the cold rolled tungsten.

Spectroscopic classification of transients

DEFF Research Database (Denmark)

Stritzinger, M. D.; Fraser, M.; Hummelmose, N. N.

2017-01-01

We report the spectroscopic classification of several transients based on observations taken with the Nordic Optical Telescope (NOT) equipped with ALFOSC, over the nights 23-25 August 2017.......We report the spectroscopic classification of several transients based on observations taken with the Nordic Optical Telescope (NOT) equipped with ALFOSC, over the nights 23-25 August 2017....

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...

Performance of neutron kinetics models for ADS transient analyses

International Nuclear Information System (INIS)

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

2002-01-01

Within the framework of the SIMMER code development, neutron kinetics models for simulating transients and hypothetical accidents in advanced reactor systems, in particular in Accelerator Driven Systems (ADSs), have been developed at FZK/IKET in cooperation with CE Cadarache. SIMMER is a fluid-dynamics/thermal-hydraulics code, coupled with a structure model and a space-, time- and energy-dependent neutronics module for analyzing transients and accidents. The advanced kinetics models have also been implemented into KIN3D, a module of the VARIANT/TGV code (stand-alone neutron kinetics) for broadening application and for testing and benchmarking. In the paper, a short review of the SIMMER and KIN3D neutron kinetics models is given. Some typical transients related to ADS perturbations are analyzed. The general models of SIMMER and KIN3D are compared with more simple techniques developed in the context of this work to get a better understanding of the specifics of transients in subcritical systems and to estimate the performance of different kinetics options. These comparisons may also help in elaborating new kinetics models and extending existing computation tools for ADS transient analyses. The traditional point-kinetics model may give rather inaccurate transient reaction rate distributions in an ADS even if the material configuration does not change significantly. This inaccuracy is not related to the problem of choosing a 'right' weighting function: the point-kinetics model with any weighting function cannot take into account pronounced flux shape variations related to possible significant changes in the criticality level or to fast beam trips. To improve the accuracy of the point-kinetics option for slow transients, we have introduced a correction factor technique. The related analyses give a better understanding of 'long-timescale' kinetics phenomena in the subcritical domain and help to evaluate the performance of the quasi-static scheme in a particular case. One

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

TASS code topical report. V.2 TASS code validation report for the non-LOCA transient analysis of the CE and Westinghouse type plants

International Nuclear Information System (INIS)

Sim, Suk K.; Chang, W. P.; Kim, K. D.; Lee, S. J.; Kim, H. C.; Yoon, H. Y.

1997-02-01

The development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the CE and Westinghouse type operating reactors as well as the PWR plants under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the FSAR transients, loss of AC power transient plant data, load rejection and startup test data for the reference plants as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best FSAR transient and shows its capability in simulating plant transient analyses. (author). 12 refs., 32 tabs., 132 figs

Studies on an Electromagnetic Transient Model of Offshore Wind Turbines and Lightning Transient Overvoltage Considering Lightning Channel Wave Impedance

Directory of Open Access Journals (Sweden)

Li Zhang

2017-12-01

Full Text Available In recent years, with the rapid development of offshore wind turbines (WTs, the problem of lightning strikes has become more and more prominent. In order to reduce the failure rate caused by the transient overvoltage of lightning struck offshore WTs, the influencing factors and the response rules of transient overvoltage are analyzed. In this paper, a new integrated electromagnetic transient model of offshore WTs is established by using the numerical calculation method of the electromagnetic field first. Then, based on the lightning model and considering the impedance of the lightning channel, the transient overvoltage of lightning is analyzed. Last, the electromagnetic transient model of offshore WTs is simulated and analyzed by using the alternative transients program electro-magnetic transient program (ATP-EMTP software. The influence factors of lightning transient overvoltage are studied. The main influencing factors include the sea depth, the blade length, the tower height, the lightning flow parameters, the lightning strike point, and the blade rotation position. The simulation results show that the influencing factors mentioned above have different effects on the lightning transient overvoltage. The results of the study have some guiding significance for the design of the lightning protection of the engine room.

Analysis of very fast transients in layer-type transformer windings

NARCIS (Netherlands)

Popov, M.; Sluis, van der L.; Smeets, R.; Lopez Roldan, J.

2007-01-01

This paper deals with the measurement, modeling, and simulation of very fast transient overvoltages in layer-type distribution transformer windings. Measurements were performed by applying a step impulse with 50-ns rise time on a single-phase test transformer equipped with measuring points along the

Control Design of VSIs to Enhance Transient Performance in Microgrids

DEFF Research Database (Denmark)

Federico, de Bosio; Antonio DeSouza Ribeiro, Luiz; Savaghebi, Mehdi

2016-01-01

This paper discusses the control design for an islanded microgrid in order to ensure acceptable performance in terms of voltage quality and load sharing by focusing on transient conditions. To this aim, state feedback decoupling approach has been applied. Experimental tests have been performed...

Future Food Production System Development Pulling From Space Biology Crop Growth Testing in Veggie

Science.gov (United States)

Massa, Gioia; Romeyn, Matt; Fritsche, Ralph

2017-01-01

Preliminary crop testing using Veggie indicates the environmental conditions provided by the ISS are generally suitable for food crop production. When plant samples were returned to Earth for analysis, their levels of nutrients were comparable to Earth-grown ground controls. Veggie-grown produce food safety microbiology analysis indicated that space-grown crops are safe to consume. Produce sanitizing wipes were used on-orbit to further reduce risk of foodborne illness. Validation growth tests indicated abiotic challenges of insufficient or excess fluid delivery, potentially reduced air flow leading to excess water, elevated CO2 leading to physiological responses, and microorganisms that became opportunistic pathogens. As NASA works to develop future space food production, several areas of research to define these systems pull from the Veggie technology validation tests. Research into effective, reusable water delivery and water recovery methods for future food production systems arises from abiotic challenges observed. Additionally, impacts of elevated CO2 and refinement of fertilizer and light recipes for crops needs to be assessed. Biotic pulls include methods or technologies to effectively sanitize produce with few consumables and low inputs; work to understand the phytomicrobiome and potentially use it to protect crops or enhance growth; selection of crops with high harvest index and desirable flavors for supplemental nutrition; crops that provide psychosocial benefits, and custom space crop development. Planning for future food production in a deep space gateway or a deep space transit vehicle requires methods of handling and storing seeds, and ensuring space seeds are free of contaminants and long-lived. Space food production systems may require mechanization and autonomous operation, with preliminary testing initiated to identify operations and capabilities that are candidates for automation. Food production design is also pulling from Veggie logistics

Fuel rod response to BWR power oscillations during anticipated transient without scram

International Nuclear Information System (INIS)

Cunningham, M.; Scott, H.

1998-01-01

The US NRC is examining fuel behaviour during a postulated BWR anticipated transient without scram (ATWS) with power oscillations to determine if current regulatory criteria are adequate. Currently, the 280 cal/g limit for RIAs is used to show that coolable geometry is maintained and pressure pulses are avoided during ATWSs. Two specific questions have now been raised about the continued use of the 280 cal/g value. First, this value was derived from energy deposition values whereas the regulatory requirements are written in terms of fuel enthalpy. The second is that fuel rod rupture with fuel dispersal has been observed in RIA tests with high bum-up fuel rods having energy deposition values well below the current limit. However, the BWR ATWS power oscillation transient is slower than a RIA power pulse, thus reducing the likelihood of failure. Therefore questions about the adequacy of the 280 cal/g limit do not necessarily imply unacceptable fuel damage occurring during such power oscillations and there is no immediate safety concern. The reported analysis, using the FRAPTRAN transient fuel rod analysis code, was thus undertaken to determine if further investigation might be appropriate and with the intention of starting some discussions about the issue. There was a comment that a limit of 100 cal/g fuel enthalpy had been mentioned following the scoping calculations but that perhaps enthalpy was not the main concern in an ATWS. It was also observed that cladding stresses are lower than in all RIA. The question was what really is the main concern. It was replied that the main concern was a question of maintaining a coolable geometry i.e. not loosing fuel particles out of the rod. And it was agreed that enthalpy may not be the important issue, rather that it previously had been used as the parameter and so had been considered. Confirmation of this presently being an evaluation and not a regulatory concern was sought and provided, it being pointed out that the NRC

Best-estimate analyses of LOFT anticipated transients with and without scram using DYNODE-P

International Nuclear Information System (INIS)

Kern, R.C.; Anderson, R.O.; Rautmann, D.A.

1984-01-01

Six LOFT transient tests with scram (L6-1, L6-2, L6-3, L6-7, L6-8B-1, and L6-8B-2) and two anticipated transient tests without scram (L9-3 and L9-4) have been analyzed using a best-estimate DYNODE-P/5.2 computer model. These tests span a wide range of anticipated operational occurrences for Pressurized Water Reactors. In general, satisfactory agreement between calculation and measurement for the key system parameters (nuclear power, primary and secondary pressures, temperatures, liquid levels, and flows) have been found. Sensitivity studies have resolved all significant discrepancies. These analyses have provided a significant qualification of the model for application to these types of events

Quantification of metabolically active transient storage (MATS) in two reaches with contrasting transient storage and ecosystem respiration

Science.gov (United States)

Alba Argerich; Roy Haggerty; Eugènia Martí; Francesc Sabater; Jay. Zarnetske

2011-01-01

Water transient storage zones are hotspots for metabolic activity in streams although the contribution of different types of transient storage zones to the whole�]reach metabolic activity is difficult to quantify. In this study we present a method to measure the fraction of the transient storage that is metabolically active (MATS) in two consecutive reaches...

Transient magnetoviscosity of dilute ferrofluids