Brazed thermocouple pass-through for sodium service in a liquid-metal-cooled fast breeder reactor

International Nuclear Information System (INIS)

Walker, D.E.

1975-10-01

Sensors installed in special fuel elements for the EBR-II reactor had 30-ft-long leads that would pass from the sodium environment through a sealed bulkhead. A hydrogen-atmosphere, induction-heated brazing furnace was constructed to simultaneously braze 20-26 separate sensor leads at one time. The brazed seals were leak-tight, and the sheath wall has less than 10 percent interaction with the braze alloy

A 3D Reconstruction Strategy of Vehicle Outline Based on Single-Pass Single-Polarization CSAR Data.

Science.gov (United States)

Leping Chen; Daoxiang An; Xiaotao Huang; Zhimin Zhou

2017-11-01

In the last few years, interest in circular synthetic aperture radar (CSAR) acquisitions has arisen as a consequence of the potential achievement of 3D reconstructions over 360° azimuth angle variation. In real-world scenarios, full 3D reconstructions of arbitrary targets need multi-pass data, which makes the processing complex, money-consuming, and time expending. In this paper, we propose a processing strategy for the 3D reconstruction of vehicle, which can avoid using multi-pass data by introducing a priori information of vehicle's shape. Besides, the proposed strategy just needs the single-pass single-polarization CSAR data to perform vehicle's 3D reconstruction, which makes the processing much more economic and efficient. First, an analysis of the distribution of attributed scattering centers from vehicle facet model is presented. And the analysis results show that a smooth and continuous basic outline of vehicle could be extracted from the peak curve of a noncoherent processing image. Second, the 3D location of vehicle roofline is inferred from layover with empirical insets of the basic outline. At last, the basic line and roofline of the vehicle are used to estimate the vehicle's 3D information and constitute the vehicle's 3D outline. The simulated and measured data processing results prove the correctness and effectiveness of our proposed strategy.

Supercritical-pressure light water cooled reactors

CERN Document Server

Oka, Yoshiaki

2014-01-01

This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

Single-pass BPM system of the Photon Factory storage ring.

Science.gov (United States)

Honda, T; Katoh, M; Mitsuhashi, T; Ueda, A; Tadano, M; Kobayashi, Y

1998-05-01

At the 2.5 GeV ring of the Photon Factory, a single-pass beam-position monitor (BPM) system is being prepared for the storage ring and the beam transport line. In the storage ring, the injected beam position during the first several turns can be measured with a single injection pulse. The BPM system has an adequate performance, useful for the commissioning of the new low-emittance lattice. Several stripline BPMs are being installed in the beam transport line. The continuous monitoring of the orbit in the beam transport line will be useful for the stabilization of the injection energy as well as the injection beam orbit.

Measurement of flow by-passing and turbulent mixing in a model of a fast-reactor steam generator

International Nuclear Information System (INIS)

Little, A.J.; Fallows, T.; Central Electricity Generating Board, Leatherhead

1989-01-01

A description is given of measurements of edge by-pass velocities and turbulent mixing in a model of a fast reactor steam generator. The velocity measurements were carried out using a DANTEC triple-split fibre probe which allowed both the speed and flow angle of a velocity vector to be measured in a plane normal to the axis of the probe. The measurements revealed the presence of reverse flows in the by-pass and adjacent in-bank channels downstream of a grid plate. The magnitude of the by-pass flow was reduced considerably by the insertion of a kicker grid at the mid point between grid plates. Turbulent mixing measurements revealed that circumferential mixing in channels near the by-pass channel was up to 5 times greater than the radial mixing. The level of radial mixing at the edge of the bank was similar to that measured near the centre of the bank. A method of transposing mass diffusion measurements in air to thermal diffusivities of sodium is discussed. (orig.)

Double Pass 595?nm pulsed dye laser at a 6 minute interval for the treatment of port-wine stains is not more effective than single pass

NARCIS (Netherlands)

Peters, M. A. D.; van Drooge, A. M.; Wolkerstorfer, A.; van Gemert, M. J. C.; van der Veen, J. P. W.; Bos, J. D.; Beek, J. F.

2012-01-01

Background Pulsed dye laser (PDL) is the first choice for treatment of port wine stains (PWS). However, outcome is highly variable and only a few patients achieve complete clearance. The objective of the study was to compare efficacy and safety of single pass PDL with double pass PDL at a 6 minute

Model for Estimation of Thermal History Produced by a Single Pass Underwater Wet Weld

National Research Council Canada - National Science Library

Dill, Jay

1997-01-01

Thermal history calculations for single pass underwater wet weldments were made by solving the appropriate beat transfer equations using the three-dimensional Crank-Nicholson finite difference method...

Small mirror fusion reactors

International Nuclear Information System (INIS)

Carlson, G.A.; Schultz, K.R.; Smith, A.C. Jr.

1978-01-01

Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

A two-stage ethanol-based biodiesel production in a packed bed reactor

DEFF Research Database (Denmark)

Xu, Yuan; Nordblad, Mathias; Woodley, John

2012-01-01

were conducted in a simulated series of reactors by repeatedly passing the reaction mixture through a single reactor, with separation of the by-product glycerol and water between passes in the first and second stages, respectively. The second stage brought the major components of biodiesel to ‘in......-spec’ levels according to the European biodiesel specifications for methanol-based biodiesel. The highest overall productivity achieved in the first stage was 2.52 kg FAEE(kg catalyst)−1 h−1 at a superficial velocity of 7.6 cm min−1, close to the efficiency of a stirred tank reactor under similar conditions...

Prospect of increases the efficiency of nuclear energy production in advanced light water reactors

International Nuclear Information System (INIS)

Proskuryakov, Konstantin; Belikov, Svyatoslav; Novikov, Konstantin

2011-01-01

Values of quality factor and pass-band for acoustic model of reactor core of reactor unit with WWER - 1000 are defined at single-phase and biphasic conditions of the coolant. The settlement-theoretical substantiation of conditions of growth of vibrations of internals and fuel assembly within a pass-band limits. Results of calculation and measurements have suitable coincidence. The technique of forecasting is developed and results of forecasting of vibration acoustical resonances of fuel assemblies with coolant in stationary and transitive operating modes of reactor unit with WWER - 1000 are worked out. (author)

Reactor instrumentation. Definition of the single failure criterion

International Nuclear Information System (INIS)

1980-12-01

The standard defines the single failure criterion which is used in other IEC publications on reactor safety systems. The purpose of the single failure criterion is the assurance of minimum redundancy. (orig./HP) [de

Nuclear reactor trip system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

Each parameter of the processes of a nuclear reactor and components operatively associated with it is monitored by a set of four like sensors. A trip system normally operates on a ''two out four'' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the ''two out of four''configuration would be reduced to a ''one out of three'' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a ''two out of three'' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor. The by-pass circuit also disables the circuit coupling the by-passed sensor to the trip circuit. (author)

Evaluation of regional pulmonary blood flow in mitral valvular heart disease using single-pass radionuclide angiocardiography

International Nuclear Information System (INIS)

Chang-Soon Koh; Byung Tae Kim; Myung Chul Lee; Bo Yeon Cho

1982-01-01

Pulmonary hypertension in mitral valvular cardiac disease has been evaluated in 122 patients by a modified upper lung/lower count ratio using single-pass radionuclide angiocardiography. The mean upper lung/lower lung radio correlates well with pulmonary artery mean (r=0.483) and wedge pressure (r=0.804). After correction surgery of the cardiac valve, the ratio decreases and returns to normal range in patients judged clinically to have good surgical benifit. This modified method using single-pass technique provides additional simple, reproducible and nontraumatic results of regional pulmonary blood flow and appears to be correlated with the degree of pulmonary hypertension in mitral heart disease

Multi-pass spectroscopic ellipsometry

International Nuclear Information System (INIS)

Stehle, Jean-Louis; Samartzis, Peter C.; Stamataki, Katerina; Piel, Jean-Philippe; Katsoprinakis, George E.; Papadakis, Vassilis; Schimowski, Xavier; Rakitzis, T. Peter; Loppinet, Benoit

2014-01-01

Spectroscopic ellipsometry is an established technique, particularly useful for thickness measurements of thin films. It measures polarization rotation after a single reflection of a beam of light on the measured substrate at a given incidence angle. In this paper, we report the development of multi-pass spectroscopic ellipsometry where the light beam reflects multiple times on the sample. We have investigated both theoretically and experimentally the effect of sample reflectivity, number of reflections (passes), angles of incidence and detector dynamic range on ellipsometric observables tanΨ and cosΔ. The multiple pass approach provides increased sensitivity to small changes in Ψ and Δ, opening the way for single measurement determination of optical thickness T, refractive index n and absorption coefficient k of thin films, a significant improvement over the existing techniques. Based on our results, we discuss the strengths, the weaknesses and possible applications of this technique. - Highlights: • We present multi-pass spectroscopic ellipsometry (MPSE), a multi-pass approach to ellipsometry. • Different detectors, samples, angles of incidence and number of passes were tested. • N passes improve polarization ratio sensitivity to the power of N. • N reflections improve phase shift sensitivity by a factor of N. • MPSE can significantly improve thickness measurements in thin films

Detection of single-copy functional genes in prokaryotic cells by two-pass TSA-FISH with polynucleotide probes.

Science.gov (United States)

Kawakami, Shuji; Hasegawa, Takuya; Imachi, Hiroyuki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi; Kubota, Kengo

2012-02-01

In situ detection of functional genes with single-cell resolution is currently of interest to microbiologists. Here, we developed a two-pass tyramide signal amplification (TSA)-fluorescence in situ hybridization (FISH) protocol with PCR-derived polynucleotide probes for the detection of single-copy genes in prokaryotic cells. The mcrA gene and the apsA gene in methanogens and sulfate-reducing bacteria, respectively, were targeted. The protocol showed bright fluorescence with a good signal-to-noise ratio and achieved a high efficiency of detection (>98%). The discrimination threshold was approximately 82-89% sequence identity. Microorganisms possessing the mcrA or apsA gene in anaerobic sludge samples were successfully detected by two-pass TSA-FISH with polynucleotide probes. The developed protocol is useful for identifying single microbial cells based on functional gene sequences. Copyright © 2011 Elsevier B.V. All rights reserved.

Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

Directory of Open Access Journals (Sweden)

Reza Akbari

2017-08-01

Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

Single DV-DXCCII Based Voltage Controlled First Order All-pass Filter with Inverting and Non-inverting responses

Directory of Open Access Journals (Sweden)

B Chaturvedi

2015-12-01

Full Text Available In this paper, a new voltage controlled first order all-pass filter is presented. The proposed circuit employs a single differential voltage dual-X second generation current conveyor (DV-DXCCII and a grounded capacitor only. The proposed all-pass filter provides both inverting and non inverting voltage-mode outputs from the same configuration simultaneously without any matching condition. Non-ideal analysis along with sensitivity analysis is also investigated. The proposed circuit has low active and passive sensitivities. As an application the proposed all-pass filter is connected in cascade to get higher order filter. The theoretical results are validated thorough PSPICE simulations using TSMC 0.18µm CMOS process parameters.

An experimental evaluation of multi-pass solar air heaters

Energy Technology Data Exchange (ETDEWEB)

Satcunanathan, S.; Persad, P.

1980-12-01

Three collectors of identical dimensions but operating in the single-pass, two-pass and three-pass modes were tested simultaneously under ambient conditions. It was found that the two-pass air heater was consistently better than the single-pass air heater over the day for the range of mass flow rates considered. It was also found that at a mass flow rate of 0.0095 kg s/sup -1/ m/sup -2/, the thermal performances of the two-pass and three-pass collectors were identical, but at higher flow rates the two-pass collector was superior to the three-pass collector, the superiority decreasing with increasing mass flow rate.

Performance Testing of Hydrodesulfurization Catalysts Using a Single-Pellet-String Reactor

NARCIS (Netherlands)

Moonen, Roel; Ras, Erik Jan; Harvey, Clare; Alles, Jeroen; Moulijn, J.A.

2017-01-01

Small-scale parallel trickle-bed reactors were used to evaluate the performance of a commercial hydrodesulfurization catalyst under industrially relevant conditions. Catalyst extrudates were loaded as a single string in reactor tubes. It is demonstrated that product sulfur levels and densities

Determination of reactor parameters by single rod experiments

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Zdravkovic, Z; Ivkovic, M; Sotic, O [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1968-10-15

The objective of this work was to determine experimentally fuel element parameters using an isolated fuel element of arbitrary construction and analyzing the accuracy of their results with the aim to apply them in analysis of reactor system. The approach is based on assumption of heterogeneous reactor theory, 'source-sink' theory. The obtained experimental results have shown the possibility of obtaining data for absorption or production properties of fuel element by analyzing the thermal and epithermal neutron density distributions around a single fuel rod placed in a sufficiently large thermal hole.

Milestone experiments for single pass UV/X-ray FELs

Science.gov (United States)

Ben-Zvi, Ilan

1995-04-01

In the past decade, significant advances have been made in the theory and technology of high brightness electron beams and single pass FELs. These developments facilitate the construction of practical UV and X-ray FELs and has prompted proposals to the DOE for the construction of such facilities. There are several important experiments to be performed before committing to the construction of dedicated user facilities. Two experiments are under construction in the IR, the UCLA self-amplified spontaneous emission experiment and the BNL laser seeded harmonic generation experiment. A multi-institution collaboration is being organized about a 210 MeV electron linac available at BNL and the 10 m long NISUS wiggler. This experiment will be done in the UV and will test various experimental aspects of electron beam dynamics, FEL exponential regime with gain guiding, start-up from noise, seeding and harmonic generation. These experiments will advance the state of FEL research and lead towards future dedicated users' facilities.

Milestone experiments for single pass UV/X-ray FELs

International Nuclear Information System (INIS)

Ben-Zvi, I.

1994-01-01

In the past decade, significant advances have been made in the theory and technology of high brightness electron beams and single pass FELS. These developments facilitate the construction of practical UV and X-ray FELs and has prompted proposals to the DOE for the construction of such facilities. There are several important experiments to be performed before committing to the construction of dedicated user facilities. Two experiments are under construction in the IR, the UCLA Self Amplified Spontaneous Emission experiment and the BNL laser seeded Harmonic Generation experiment. A multi-institution collaboration is being organized about a 210 MeV electron linac available at BNL and the 10 meter tong NISUS wiggler. This experiment will be done in the UV and will test various experimental aspects of electron beam dynamics, FEL exponential regime with gain guiding, start up from noise, seeding and harmonic generation. These experiments will advance the state of FEL research and lead towards future dedicated users' facilities

Single-Pass Percutaneous Liver Biopsy for Diffuse Liver Disease Using an Automated Device: Experience in 154 Procedures

International Nuclear Information System (INIS)

Rivera-Sanfeliz, Gerant; Kinney, Thomas B.; Rose, Steven C.; Agha, Ayad K.M.; Valji, Karim; Miller, Franklin J.; Roberts, Anne C.

2005-01-01

Purpose: To describe our experience with ultrasound (US)-guided percutaneous liver biopsies using the INRAD 18G Express core needle biopsy system.Methods: One hundred and fifty-four consecutive percutaneous core liver biopsy procedures were performed in 153 men in a single institution over 37 months. The medical charts, pathology reports, and radiology files were retrospectively reviewed. The number of needle passes, type of guidance, change in hematocrit level, and adequacy of specimens for histologic analysis were evaluated.Results: All biopsies were performed for histologic staging of chronic liver diseases. The majority of patients had hepatitis C (134/153, 90.2%). All patients were discharged to home after 4 hr of postprocedural observation. In 145 of 154 (94%) biopsies, a single needle pass was sufficient for diagnosis. US guidance was utilized in all but one of the procedures (153/154, 99.4%). The mean hematocrit decrease was 1.2% (44.1-42.9%). Pain requiring narcotic analgesia, the most frequent complication, occurred in 28 of 154 procedures (18.2%). No major complications occurred. The specimens were diagnostic in 152 of 154 procedures (98.7%).Conclusions: Single-pass percutaneous US-guided liver biopsy with the INRAD 18G Express core needle biopsy system is safe and provides definitive pathologic diagnosis of chronic liver disease. It can be performed on an outpatient basis. Routine post-biopsy monitoring of hematocrit level in stable, asymptomatic patients is probably not warranted

Heat removing device for reactor container

International Nuclear Information System (INIS)

Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

1996-01-01

A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

Pressurized water reactor flow arrangement

International Nuclear Information System (INIS)

Gibbons, J.F.; Knapp, R.W.

1980-01-01

A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

Atrial electrogram quality in single-pass defibrillator leads with floating atrial bipole in patients with permanent atrial fibrillation and cardiac resynchronization therapy.

Science.gov (United States)

Sticherling, Christian; Müller, Dirk; Schaer, Beat A; Krüger, Silke; Kolb, Christof

2018-03-27

Many patients receiving cardiac resynchronization therapy (CRT) suffer from permanent atrial fibrillation (AF). Knowledge of the atrial rhythm is important to direct pharmacological or interventional treatment as well as maintaining AV-synchronous biventricular pacing if sinus rhythm can be restored. A single pass single-coil defibrillator lead with a floating atrial bipole has been shown to obtain reliable information about the atrial rhythm but has never been employed in a CRT-system. The purpose of this study was to assess the feasibility of implanting a single coil right ventricular ICD lead with a floating atrial bipole and the signal quality of atrial electrograms (AEGM) in CRT-defibrillator recipients with permanent AF. Seventeen patients (16 males, mean age 73 ± 6 years, mean EF 25 ± 5%) with permanent AF and an indication for CRT-defibrillator placement were implanted with a designated CRT-D system comprising a single pass defibrillator lead with a atrial floating bipole. They were followed-up for 103 ± 22 days using remote monitoring for AEGM transmission. All patients had at last one AEGM suitable for atrial rhythm diagnosis and of 100 AEGM 99% were suitable for visual atrial rhythm assessment. Four patients were discharged in sinus rhythm and one reverted to AF during follow-up. Atrial electrograms retrieved from a single-pass defibrillator lead with a floating atrial bipole can be reliably used for atrial rhythm diagnosis in CRT recipients with permanent AF. Hence, a single pass ventricular defibrillator lead with a floating bipole can be considered in this population. Copyright © 2018 Indian Heart Rhythm Society. Production and hosting by Elsevier B.V. All rights reserved.

High aspect ratio catalytic reactor and catalyst inserts therefor

Science.gov (United States)

Lin, Jiefeng; Kelly, Sean M.

2018-04-10

The present invention relates to high efficient tubular catalytic steam reforming reactor configured from about 0.2 inch to about 2 inch inside diameter high temperature metal alloy tube or pipe and loaded with a plurality of rolled catalyst inserts comprising metallic monoliths. The catalyst insert substrate is formed from a single metal foil without a central supporting structure in the form of a spiral monolith. The single metal foil is treated to have 3-dimensional surface features that provide mechanical support and establish open gas channels between each of the rolled layers. This unique geometry accelerates gas mixing and heat transfer and provides a high catalytic active surface area. The small diameter, high aspect ratio tubular catalytic steam reforming reactors loaded with rolled catalyst inserts can be arranged in a multi-pass non-vertical parallel configuration thermally coupled with a heat source to carry out steam reforming of hydrocarbon-containing feeds. The rolled catalyst inserts are self-supported on the reactor wall and enable efficient heat transfer from the reactor wall to the reactor interior, and lower pressure drop than known particulate catalysts. The heat source can be oxygen transport membrane reactors.

Double Pass 595 nm Pulsed Dye Laser Does Not Enhance the Efficacy of Port Wine Stains Compared with Single Pass: A Randomized Comparison with Histological Examination.

Science.gov (United States)

Yu, Wenxin; Zhu, Jiafang; Wang, Lizhen; Qiu, Yajing; Chen, Yijie; Yang, Xi; Chang, Lei; Ma, Gang; Lin, Xiaoxi

2018-03-27

To compare the efficacy and safety of double-pass pulsed dye laser (DWL) and single-pass PDL (SWL) in treating virgin port wine stain (PWS). The increase in the extent of vascular damage attributed to the use of double-pass techniques for PWS remains inconclusive. A prospective, side-by-side comparison with a histological study for virgin PWS is still lacking. Twenty-one patients (11 flat PWS, 10 hypertrophic PWS) with untreated PWS underwent 3 treatments at 2-month intervals. Each PWS was divided into three treatment sites: SWL, DWL, and untreated control. Chromametric and visual evaluation of the efficacy and evaluation of side effects were conducted 3 months after final treatment. Biopsies were taken at the treated sites immediately posttreatment. Chromametric and visual evaluation suggested that DWL sites showed no significant improvement compared with SWL (p > 0.05) in treating PWS. The mean depth of photothermal damage to the vessels was limited to a maximum of 0.36-0.41 mm in both SWL and DWL sides. Permanent side effects were not observed in any patients. Double-pass PDL does not enhance PWS clearance. To improve the clearance of PWS lesions, either the depth of laser penetration should be increased or greater photothermal damage to vessels should be generated.

Photoacoustic Soot Spectrometer (PASS) Instrument Handbook

Energy Technology Data Exchange (ETDEWEB)

Dubey, M [Los Alamos National Laboratory; Springston, S [Brookhaven National Laboratory; Koontz, A [Pacific Northwest National Laboratory; Aiken, A [Los Alamos National Laboratory

2013-01-17

The photoacoustic soot spectrometer (PASS) measures light absorption by aerosol particles. As the particles pass through a laser beam, the absorbed energy heats the particles and in turn the surrounding air, which sets off a pressure wave that can be detected by a microphone. The PASS instruments deployed by ARM can also simultaneously measure the scattered laser light at three wavelengths and therefore provide a direct measure of the single-scattering albedo. The Operator Manual for the PASS-3100 is included here with the permission of Droplet Measurement Technologies, the instrument’s manufacturer.

Temperature fluctuation reducing device for FBR type reactor

International Nuclear Information System (INIS)

Ootsuka, Fumio; Shiratori, Fumihiro.

1991-01-01

In existent FBR type reactors, since temperature fluctuation in the reactor upper portion has been inevitable, thermal fatigue may be caused possibly in reactor core upper mechanisms. Then, a valve is disposed to a control rod lower guide tube contained in a reactor container for automatically controlling the amount of passing coolants in accordance with the temperature of the passing coolants, to mix and control coolants passing through a fuel assembly in adjacent with the guide tube and coolants passing through the guide tube. Further, a rectification cylinder is disposed, in which a portion of coolants passing through the fuel assembly is caused to flow. An orifice is disposed to the cylinder with an exit being disposed to the upstream thereof such that the coolants not flown into the rectification cylinder and the coolants passing through the guide tube are mixed to moderate the temperature fluctuation. That is, a portion of the coolants flown into the rectification cylinder can not pass through the orifice, but flow backwardly to the upstream and is discharged out of the rectification cylinder from the coolants exit and mixed sufficiently with coolants passing through the guide tube. In this way, temperature fluctuation can be moderated. (N.H.)

Static thermo-optic instability in double-pass fiber amplifiers

DEFF Research Database (Denmark)

Lægsgaard, Jesper

2016-01-01

A coupled-mode formalism, earlier used to describe transverse mode instabilities in single-pass optical fiber amplifiers, is extended to the case of double-pass amplifiers. Contrary to the single-pass case, it is shown that the thermo-optic nonlinearity can couple light at the same frequency...... between the LP01 and LP11 modes, leading to a static deformation of the output beam profile. This novel phenomenon is caused by the interaction of light propagating in either direction with thermo-optic index perturbations caused by light propagating in the opposite direction. The threshold power...... for the static deformation is found to be several times lower than what is typically found for the dynamic modal instabilities observed in single-pass amplifiers. (C) 2016 Optical Society of America...

Reactor operation method

International Nuclear Information System (INIS)

Osumi, Katsumi; Miki, Minoru.

1979-01-01

Purpose: To prevent stress corrosion cracks by decreasing the dissolved oxygen and hydrogen peroxide concentrations in the coolants within a reactor container upon transient operation such as at the start-up or shutdown of bwr type reactors. Method: After a condensate has been evacuated, deaeration operation is conducted while opening a main steam drain line, as well as a main steam separation valve and a by-pass valve in a turbine by-pass line connecting the main steam line and the condenser without by way of a turbine, and the reactor is started-up by the extraction of control rods after the concentration of dissolved oxygen in the cooling water within a pressure vessel has been decreased below a predetermined value. Nuclear heating is started after the reactor water has been increased to about 150 0 C by pump heating after the end of the deaeration operation for preventing the concentration of hydrogen peroxide and oxygen in the reactor water from temporarily increasing immediately after the start-up. The corrosive atmosphere in the reactor vessel can thus be moderated. (Horiuchi, T.)

Optical performance of multifocal soft contact lenses via a single-pass method.

Science.gov (United States)

Bakaraju, Ravi C; Ehrmann, Klaus; Falk, Darrin; Ho, Arthur; Papas, Eric

2012-08-01

A physical model eye capable of carrying soft contact lenses (CLs) was used as a platform to evaluate optical performance of several commercial multifocals (MFCLs) with high- and low-add powers and a single-vision control. Optical performance was evaluated at three pupil sizes, six target vergences, and five CL-correcting positions using a spatially filtered monochromatic (632.8 nm) light source. The various target vergences were achieved by using negative trial lenses. A photosensor in the retinal plane recorded the image point-spread that enabled the computation of visual Strehl ratios. The centration of CLs was monitored by an additional integrated en face camera. Hydration of the correcting lens was maintained using a humidity chamber and repeated instillations of rewetting saline drops. All the MFCLs reduced performance for distance but considerably improved performance along the range of distance to near target vergences, relative to the single-vision CL. Performance was dependent on add power, design, pupil, and centration of the correcting CLs. Proclear (D) design produced good performance for intermediate vision, whereas Proclear (N) design performed well at near vision (p 4 mm in diameter. Acuvue Oasys bifocal produced performance comparable with single-vision CL for most vergences. Direct measurement of single-pass images at the retinal plane of a physical model eye used in conjunction with various MFCLs is demonstrated. This method may have utility in evaluating the relative effectiveness of commercial and prototype designs.

Generation of coherent soft x-rays using a single-pass free-electron laser amplifier

International Nuclear Information System (INIS)

Wang, T.F.; Goldstein, J.C.; Newnam, B.E.; McVey, B.D.

1988-01-01

We consider a single-pass free-electron laser (FEL) amplifier, driven by an rf-linac followed by a damping ring for reduced emittance, for use in generating coherent light in the soft x-ray region. The dependence of the optical gain on electron-beam quality, studied with the three-dimensional FEL simulation code FELEX, is given and related to the expected power of self-amplified spontaneous emission. We discuss issues for the damping ring designed to achieve the required electron beam quality. The idea of a multipass regenerative amplifier is also presented

Reactors set for mini market

International Nuclear Information System (INIS)

Knox, Richard.

1988-01-01

Commercial nuclear power generation on a large-scale has an uncertain future. However, it is hoped that a small nuclear reactor could form the basis for providing heating, cooling or electricity in large buildings. Based on the Slowpoke research reactor, the Slowpoke energy system concept is simple. The concept and the way in which the small-scale reactor would work are discussed. The system consists of a stainless steel tank surrounded by reinforced concrete and let into the ground. The tank is full of light water which is heated to about 90 deg C by a central core of 2.4 percent enriched uranium fuel. The resulting natural circulation causes the water to pass through a heat exchanger. The water from the heat exchanger can be used for building or district heating, to operate air-conditioners or to generate small quantities of electricity. It is hoped to automate the operation of the reactor so that continuous supervision by a team of engineers would be unnecessary. A single operator on call in the building would be able to take control actions if the reactor's safety system failed. (UK)

Development of a simultaneous partial nitrification and anaerobic ammonia oxidation process in a single reactor.

Science.gov (United States)

Cho, Sunja; Fujii, Naoki; Lee, Taeho; Okabe, Satoshi

2011-01-01

Up-flow oxygen-controlled biofilm reactors equipped with a non-woven fabric support were used as a single reactor system for autotrophic nitrogen removal based on a combined partial nitrification and anaerobic ammonium oxidation (anammox) reaction. The up-flow biofilm reactors were initiated as either a partial nitrifying reactor or an anammox reactor, respectively, and simultaneous partial nitrification and anammox was established by careful control of the aeration rate. The combined partial nitrification and anammox reaction was successfully developed in both biofilm reactors without additional biomass inoculation. The reactor initiated as the anammox reactor gave a slightly higher and more stable mean nitrogen removal rate of 0.35 (±0.19) kg-N m(-3) d(-1) than the reactor initiated as the partial nitrifying reactor (0.23 (±0.16) kg-N m(-3) d(-1)). FISH analysis revealed that the biofilm in the reactor started as the anammox reactor were composed of anammox bacteria located in inner anoxic layers that were surrounded by surface aerobic AOB layers, whereas AOB and anammox bacteria were mixed without a distinguishable niche in the biofilm in the reactor started as the partial nitrifying reactor. However, it was difficult to efficiently maintain the stable partial nitrification owing to inefficient aeration in the reactor, which is a key to development of the combined partial nitrification and anammox reaction in a single biofilm reactor. Copyright © 2010 Elsevier Ltd. All rights reserved.

Nitrite reduction and methanogenesis in a single-stage UASB reactor.

Science.gov (United States)

Borges, L I; López-Vazquez, C M; García, H; van Lier, J B

2015-01-01

In this study, nitrite reduction and methanogenesis in a single-stage upflow anaerobic sludge blanket (UASB) reactor was investigated, using high-strength synthetic domestic wastewater as substrate. To assess long-term effects and evaluate the mechanisms that allow successful nitrite reduction and methanogenesis in a single-stage UASB, sludge was exposed to relatively high nitrite loading rates (315 ± 13 mgNO(2)(-)-N/(l.d)), using a chemical oxygen demand (COD) to nitrogen ratio of 18 gCOD/gNO(2)(-)-N, and an organic loading rate of 5.4 ± 0.2 gCOD/(l.d). In parallel, the effects of sludge morphology on methanogenesis inhibition were studied by performing short-term batch activity tests at different COD/NO(2)(-)-N ratios with anaerobic sludge samples. In long-term tests, denitrification was practically complete and COD removal efficiency did not change significantly after nitrite addition. Furthermore, methane production only decreased by 13%, agreeing with the reducing equivalents requirement for complete NO(2)(-) reduction to N₂. Apparently, the spatial separation of denitrification and methanogenesis zones inside the UASB reactor allowed nitrite reduction and methanogenesis to occur at the same moment. Batch tests showed that granules seem to protect methanogens from nitrite inhibition, probably due to transport limitations. Combined COD and N removal via nitrite in a single-stage UASB reactor could be a feasible technology to treat high-strength domestic wastewater.

Studies on water turbine runner which fish can pass through: In case of single stage axial runner

International Nuclear Information System (INIS)

Shimizu, Yukimari; Maeda, Takao; Nagoshi, Osamu; Ieda, Kazuma; Shinma, Hisako; Hagimoto, Michiko

1994-01-01

The relationship between water turbine runner design and operation and the safe passage of fish through the turbine is studied. The kinds of fish used in the tests are a dace, a sweet fish and a small salmon. A single stage axial runner is used. The velocity and pressure distributions were measured inside the turbine casing and along the casing wall. Many pictures showing fish passing through the rotating runner were taken and analyzed. The swimming speed of the fish was examined from video recordings. Fish pass through the runner more rapidly when they can determine and choose the easier path. Injury and mortality of fish are affected by the runner speed and the location of impact of the runner on the fish body

Coordinate control of integral reactor based on single neuron PID controller

International Nuclear Information System (INIS)

Liu Yan; Xia Hong

2014-01-01

As one of the main type of reactors in the future, the development of the integral reactor has attracted worldwide attention. On the basis of understanding the background of the integral reactor, the author will be familiar with and master the power control of reactor and the feedwater flow control of steam generator, and the speed control of turbine (turbine speed control is associated with the turbine load control). According to the expectative program 'reactor power following turbine load' of the reactor, it will make coordinate control of the three and come to a overall control scheme. The author will use the supervisory learning algorithm of Hebb for single neuron PID controller with self-adaptation to study the coordinate control of integral reactor. Compared with conventional PI or PID controller, to a certain extent, it solves the problems that traditional PID controller is not easy to tune real-time parameters and lack of effective control for a number of complex processes and slow-varying parameter systems. It improves the security, reliability, stability and flexibility of control process and achieves effective control of the system. (authors)

Experimental study of the core grid by-pass orifices inlet pressure drop of the new core of the R A 6 reactor

International Nuclear Information System (INIS)

Masson, V. P; Garcia, J. C; Delmastro, D. F

2006-01-01

In this work the core grid by-pass orifices inlet pressure drop of the new core of the R A6 reactor are experimentally studied.The experiments are performed using a 1:1 scale mock-up of an external fuel element cell.Different gaps between fuel elements are considered in order to take into account the design allowances. Different flows are considered to take into account the normal operation flow range.Measurement uncertainties are included.The results will be used to calculate the core flow distribution [es

High-Accuracy Elevation Data at Large Scales from Airborne Single-Pass SAR Interferometry

Directory of Open Access Journals (Sweden)

Guy Jean-Pierre Schumann

2016-01-01

Full Text Available Digital elevation models (DEMs are essential data sets for disaster risk management and humanitarian relief services as well as many environmental process models. At present, on the hand, globally available DEMs only meet the basic requirements and for many services and modeling studies are not of high enough spatial resolution and lack accuracy in the vertical. On the other hand, LiDAR-DEMs are of very high spatial resolution and great vertical accuracy but acquisition operations can be very costly for spatial scales larger than a couple of hundred square km and also have severe limitations in wetland areas and under cloudy and rainy conditions. The ideal situation would thus be to have a DEM technology that allows larger spatial coverage than LiDAR but without compromising resolution and vertical accuracy and still performing under some adverse weather conditions and at a reasonable cost. In this paper, we present a novel single pass In-SAR technology for airborne vehicles that is cost-effective and can generate DEMs with a vertical error of around 0.3 m for an average spatial resolution of 3 m. To demonstrate this capability, we compare a sample single-pass In-SAR Ka-band DEM of the California Central Valley from the NASA/JPL airborne GLISTIN-A to a high-resolution LiDAR DEM. We also perform a simple sensitivity analysis to floodplain inundation. Based on the findings of our analysis, we argue that this type of technology can and should be used to replace large regions of globally available lower resolution DEMs, particularly in coastal, delta and floodplain areas where a high number of assets, habitats and lives are at risk from natural disasters. We conclude with a discussion on requirements, advantages and caveats in terms of instrument and data processing.

A Novel Single Pass Authenticated Encryption Stream Cipher for Software Defined Radios

DEFF Research Database (Denmark)

Khajuria, Samant

2012-01-01

to propose cryptographic services such as confidentiality, integrity and authentication. Therefore, integration of security services into SDR devices is essential. Authenticated Encryption schemes donate the class of cryptographic algorithms that are designed for protecting both message confidentiality....... This makes authenticated encryption very attractive for low-cost low-power hardware implementations, as it allows for the substantial decrease in the circuit area and power consumed compared to the traditional schemes. In this thesis, an authenticated encryption scheme is proposed with the focus of achieving...... high throughput and low overhead for SDRs. The thesis is divided into two research topics. One topic is the design of a 1-pass authenticated encryption scheme that can accomplish both message secrecy and authenticity in a single cryptographic primitive. The other topic is the implementation...

Feasibility of a single-purpose reactor plant for district heating in Finland

International Nuclear Information System (INIS)

Tarjanne, R.; Vuori, S.; Eerikaeinen, L.; Saukkoriipi, L.

A feasibility study of a single-purpose reactor for district heating is presented. The reactor chosen is of an ordinary pressurized water reactor type with a thermal output of 100 to 200 MW. Primary circuit steam generators employed in ordinary PWR's are replaced by water-water heat exchangers. For safety reasons an intermediate circuit separates the primary from the network water. The conditions of the district heating systems in Finland were taken into account, which led to the choice of an average temperature of 160 0 C for the reactor coolant and a pressure of 13.5 bar. This, coupled with minimal control requirements helped design a considerably simple reactor plant. On condition, the reactor satisfies the basic heat demand in a district heating system, the effective annual full-power operation time of the heating reactor is from 5000 h to 7000 h. Economic comparisons indicated that the heating reactor may be competitive if the operation period is of this order. As the reactor has to be sited near the heat consumption area for reasons of economy, the safety aspects are of major importance and may in themselves preclude the realization of the heating idea. (author)

Isolation colling device for reactor

International Nuclear Information System (INIS)

Ikehara, Morihiko; Arai, Shigeki.

1982-01-01

Purpose: To prevent undesired operation of an emergency core cooling system due to excess lowering of water level in a reactor. Constitution: In an emergency facility adapted to drive a turbine, upon reactor isolation, with the excess steams of the reactor to operate a pump and thereby inject cooling water to the reactor, a water level detector is provided and connected to a pump exhaust valve control circuit, a turbine inlet valve control circuit and a by-pass valve control circuit. Valve ON-OFF is automatically controlled depending on the water level to thereby render the level constant. A by-pass pipe is branched from a pump exhaust pipe and connected to a condensate storage tank. When the water level rises due to water injection, the injecting water is returned to circulate by way of the by-pass pipe to the condensate storage tank under the ON-OFF for each of the valves while the turbine being kept to drive. Then, if the water level is lowered, water injection is started again by the ON-OFF for each of the valves. (Ikeda, J.)

Effects of multi-pass arc welding on mechanical properties of carbon steel C25 plate

International Nuclear Information System (INIS)

Adedayo, S.M.; Babatunde, A.S.

2013-01-01

The effects of multi-pass welding on mechanical properties of C25 carbon steel plate were examined. Mild steel plate workpieces of 90 x 55 mm 2 area and 10 mm thickness with a 30 degrees vee weld-grooves were subjected to single and multi-pass welding. Toughness, hardness and tensile tests of single and multi-pass welds were conducted. Toughness values of the welds under double pass welds were higher than both single pass and unwelded alloy, at respective maximum values of 2464, 2342 and 2170 kN/m. Hardness values were reduced under double pass relative to single pass welding with both being lower than the value for unwelded alloy; the values were 40.5, 43.2 and 48.5 Rs respectively at 12 mm from the weld line. The tensile strength of 347 N/mm 2 under multi-pass weld was higher than single pass weld with value of 314 N/mm 2 . Therefore, the temperature distribution and apparent pre-heating during multi-pass welding increased the toughness and tensile strength of the weldments, but reduced the hardness. (au)

Generic, network schema agnostic sparse tensor factorization for single-pass clustering of heterogeneous information networks.

Science.gov (United States)

Wu, Jibing; Meng, Qinggang; Deng, Su; Huang, Hongbin; Wu, Yahui; Badii, Atta

2017-01-01

Heterogeneous information networks (e.g. bibliographic networks and social media networks) that consist of multiple interconnected objects are ubiquitous. Clustering analysis is an effective method to understand the semantic information and interpretable structure of the heterogeneous information networks, and it has attracted the attention of many researchers in recent years. However, most studies assume that heterogeneous information networks usually follow some simple schemas, such as bi-typed networks or star network schema, and they can only cluster one type of object in the network each time. In this paper, a novel clustering framework is proposed based on sparse tensor factorization for heterogeneous information networks, which can cluster multiple types of objects simultaneously in a single pass without any network schema information. The types of objects and the relations between them in the heterogeneous information networks are modeled as a sparse tensor. The clustering issue is modeled as an optimization problem, which is similar to the well-known Tucker decomposition. Then, an Alternating Least Squares (ALS) algorithm and a feasible initialization method are proposed to solve the optimization problem. Based on the tensor factorization, we simultaneously partition different types of objects into different clusters. The experimental results on both synthetic and real-world datasets have demonstrated that our proposed clustering framework, STFClus, can model heterogeneous information networks efficiently and can outperform state-of-the-art clustering algorithms as a generally applicable single-pass clustering method for heterogeneous network which is network schema agnostic.

Space nuclear reactor concepts for avoidance of a single point failure

International Nuclear Information System (INIS)

El-Genk, M. S.

2007-01-01

This paper presents three space nuclear reactor concepts for future exploration missions requiring electrical power of 10's to 100's kW, for 7-10 years. These concepts avoid a single point failure in reactor cooling; and they could be used with a host of energy conversion technologies. The first is lithium or sodium heat pipes cooled reactor. The heat pipes operate at a fraction of their prevailing capillary or sonic limit. Thus, when a number of heat pipes fail, those in the adjacent modules remove their heat load, maintaining reactor core adequately cooled. The second is a reactor with a circulating liquid metal coolant. The reactor core is divided into six identical sectors, each with a separate energy conversion loop. The sectors in the reactor core are neurotically coupled, but hydraulically decoupled. Thus, when a sector experiences a loss of coolant, the fission power generated in it will be removed by the circulating coolant in the adjacent sectors. In this case, however, the reactor fission power would have to decrease to avoid exceeding the design temperature limits in the sector with a failed loop. These two reactor concepts are used with energy conversion technologies, such as advanced Thermoelectric (TE), Free Piston Stirling Engines (FPSE), and Alkali Metal Thermal-to- Electric Conversion (AMTEC). Gas cooled reactors are a better choice to use with Closed Brayton Cycle engines, such as the third reactor concept to be presented in the paper. It has a sectored core that is cooled with a binary mixture of He-Xe (40 gm/mole). Each of the three sectors in the reactor has its own CBC and neutronically, but not hydraulically, coupled to the other sectors

Siting analysis and risk assessment for small single-purpose heating reactors

International Nuclear Information System (INIS)

Tarjanne, R.

1979-04-01

Two alternative sites both 10km away from the centre of Helsinki are considered for reactor unit sizes of 400mw and 800mw. The risks associated with a small single-purpose heating reactor is evaluated for normal operation and accident conditions. The evaluation for accident condition is performed for three characteristics accidents. Three pathways are considered in the calculation of the radiation exposure: direct external gamma dose from the release plume, direct gamma radiation from deposited activity on the ground and dose due to inhalation. The risks are compared with the risks from alternative conventional fossil fuelled district heat production methods. The results show that the heating reactor alternative causes an unsignificant risk, which is far less than the risk caused by the fossil-fuelled alternatives

Nuclear reactor power supply system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

Cooling nuclear reactor fuel

International Nuclear Information System (INIS)

Porter, W.H.L.

1975-01-01

Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

Continuous-wave sodium D2 resonance radiation generated in single-pass sum-frequency generation with periodically poled lithium niobate.

Science.gov (United States)

Yue, J; She, C-Y; Williams, B P; Vance, J D; Acott, P E; Kawahara, T D

2009-04-01

With two cw single-mode Nd:YAG lasers at 1064 and 1319 nm and a periodically poled lithium niobate crystal, 11 mW of 2 kHz/100 ms bandwidth single-mode tunable 589 nm cw radiation has been detected using single-pass sum-frequency generation. The demonstrated conversion efficiency is approximately 3.2%[W(-1) cm(-1)]. This compact solid-state light source has been used in a solid-state-dye laser hybrid sodium fluorescence lidar transmitter to measure temperatures and winds in the upper atmosphere (80-105 km); it is being implemented into the transmitter of a mobile all-solid-state sodium temperature and wind lidar under construction.

Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

Energy Technology Data Exchange (ETDEWEB)

Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

2015-06-15

Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

International Nuclear Information System (INIS)

Labib, Satira; King, Jeffrey

2015-01-01

Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort

Energy production from distillery wastewater using single and double-phase upflow anaerobic sludge blanket (UASB) reactor

Energy Technology Data Exchange (ETDEWEB)

Muyodi, F J; Rubindamayugi, M S.T. [Univ. of Dar es Salaam, Applied Microbiology Unit (Tanzania, United Republic of)

1998-12-31

A Single-phase (SP) and Double-phase (DP) Upflow Anaerobic Sludge Blanket (UASB) reactors treating distillery wastewater were operated in parallel. The DP UASB reactor showed better performance than the SP UASB reactor in terms of maximum methane production rate, methane content and Chemical Oxygen Demand (COD) removal efficiency. (au) 20 refs.

Energy production from distillery wastewater using single and double-phase upflow anaerobic sludge blanket (UASB) reactor

Energy Technology Data Exchange (ETDEWEB)

Muyodi, F.J.; Rubindamayugi, M.S.T. [Univ. of Dar es Salaam, Applied Microbiology Unit (Tanzania, United Republic of)

1997-12-31

A Single-phase (SP) and Double-phase (DP) Upflow Anaerobic Sludge Blanket (UASB) reactors treating distillery wastewater were operated in parallel. The DP UASB reactor showed better performance than the SP UASB reactor in terms of maximum methane production rate, methane content and Chemical Oxygen Demand (COD) removal efficiency. (au) 20 refs.

Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

Science.gov (United States)

Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

2017-07-01

In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

The Spectral Shift Control Reactor as an option for much improved uranium utilisation in single-batch SMRs

Energy Technology Data Exchange (ETDEWEB)

Lindley, B.A., E-mail: bal29@cam.ac.uk; Parks, G.T.

2016-12-01

Highlights: • A PWR with mixed D{sub 2}O/H{sub 2}O moderator/coolant is investigated for SMR applications. • Heavy water concentration varied over the cycle to give ‘spectral shift’ operation. • Much wetter lattice than normal is neutronically favourable. • Taller fuel stack is thus needed to ensure acceptable MDNBR. • 35–43% increase in uranium utilisation for single batch reactor is possible. - Abstract: The Spectral Shift Control Reactor (SSCR) uses a mix of D{sub 2}O and H{sub 2}O to moderate and cool the reactor. Initially, a high proportion of D{sub 2}O is used, such that the reactor is substantially under-moderated, with excess neutrons being primarily captured in {sup 238}U, breeding {sup 239}Pu. Towards the end of the cycle (EOC), the coolant is predominantly H{sub 2}O, thermalising the neutron spectrum and increasing reactivity. Recently, small modular reactors (SMRs) have gained significant interest as a means of providing a power source that requires little maintenance and refuelling. This motivates long cycles and reduced batch operation. For a single-batch reactor, there is typically a 33% penalty to uranium utilisation compared to a 3-batch reactor. Lattice calculations demonstrate the potential of the SSCR to greatly improve uranium utilisation in single-batch reactors over a range of enrichments. A relatively ‘wet’ lattice is employed which further improves uranium utilisation. Cases with 5% and 15% fissile loading are considered, for which it is respectively possible to achieve 47% and 39% increases in natural uranium utilisation using the SSCR relative to a ‘reference’ light water reactor. In the latter case, if 25% thorium is mixed into the fuel, the improvement in uranium utilisation increases to a total of 49%. Hence, in both cases, it is possible to in effect eliminate the penalty of using a single fuel batch. The ‘wet’ lattice introduces substantial thermal-hydraulic challenges due to the significantly higher fuel

Experimental and simulation studies on a single pass, double duct solar air heater

Energy Technology Data Exchange (ETDEWEB)

Forson, F.K. [Kwame Nkrumah Univ. of Science and Technology, Dept. of Mechanical Engineering, Kumasi (Ghana); Rajakaruna, H. [De Montfort Univ., School of Engineering and Technology, Leicester (United Kingdom)

2003-05-01

A mathematical model of a single pass, double duct solar air heater (SPDDSAH) is described. The model provides a design tool capable of predicting: incident solar radiation, heat transfer coefficients, mean air flow rates, mean air temperature and relative humidity at the exit. Results from the simulation are presented and compared with experimental ones obtained on a full scale air heater and a small scale laboratory one. Reasonable agreement between the predicted and measured values is demonstrated. Predicted results from a parametric study are also presented. It is shown that significant improvement in the SPDDSAH performance can be obtained with an appropriate choice of the collector parameters and the top to bottom channel depth ratio of the two ducts. The air mass flow rate is shown to be the dominant factor in determining the overall efficiency of the heater. (Author)

A Unified Algorithm for Channel Imbalance and Antenna Phase Center Position Calibration of a Single-Pass Multi-Baseline TomoSAR System

Directory of Open Access Journals (Sweden)

Yuncheng Bu

2018-03-01

Full Text Available The multi-baseline synthetic aperture radar (SAR tomography (TomoSAR system is employed in such applications as disaster remote sensing, urban 3-D reconstruction, and forest carbon storage estimation. This is because of its 3-D imaging capability in a single-pass platform. However, a high 3-D resolution of TomoSAR is based on the premise that the channel imbalance and antenna phase center (APC position are precisely known. If this is not the case, the 3-D resolution performance will be seriously degraded. In this paper, a unified algorithm for channel imbalance and APC position calibration of a single-pass multi-baseline TomoSAR system is proposed. Based on the maximum likelihood method, as well as the least squares and the damped Newton method, we can calibrate the channel imbalance and APC position. The algorithm is suitable for near-field conditions, and no phase unwrapping operation is required. The effectiveness of the proposed algorithm has been verified by simulation and experimental results.

Mechanical design of core components for a high performance light water reactor with a three pass core

International Nuclear Information System (INIS)

Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

2007-01-01

Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

Fuel assembly outlet temperature profile influence on core by-pass flow and power distribution determination in WWER -440 reactors

International Nuclear Information System (INIS)

Petenyi, V.; Klucarova, K.; Remis, J.

2003-01-01

The in core instrumentation of the WWER-440 reactors consists of the thermocouple system and the system of self powered detectors (SPD). The thermocouple systems are positioned about 50 cm above the fuel bundle upper flow-mixing grid. The usual assumption is that, the coolant is well mixed in the Tc location, i.e. the temperature is constant through the flow cross-section area. The present evaluations by using the FLUENT 5.5.14 code reveal that, this assumption is not fulfilled. There exists a temperature profile that depends on fuel assembly geometry and on inner power profile of the fuel assembly. The paper presents the estimation of this effect and its influence on the core power distribution and the core by-pass flow determination. Comparison with measurements in Mochovce NPP will also be a part of this presentation (Authors)

Three-dimensional two-fluid numerical treatment of a reactor vessel in TRAC

International Nuclear Information System (INIS)

Liles, D.R.

1979-01-01

A three-dimensional two-fluid finite difference model has been used in TRAC (Transient Reactor Analysis Code) to represent a pressurized water reactor vessel. Mesh cells may be blocked off completely to represent large flow obstructions such as downcomer walls. The hydrodynamic volumes and flow areas may also be reduced in order to provide a porous matrix simulation of smaller scale strucuture. The finite difference equations are semi-implicit so that stability time scales are associated with material movement and not wave propagation. The block matrix structure is reduced during the implicit pass to a single element seven stripe system which is easily solved iteratively. This procedure has successfully performed numerous simulations of both full sized reactor accidents and smaller scale experments. It has proven to be a useful feature of the TRAC effort

Pressure tube reactors

International Nuclear Information System (INIS)

Natori, Hisahide.

1981-01-01

Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

Effect of a ballast zone on the hydraulic stability of a single-pass steam generator

International Nuclear Information System (INIS)

Belyakov, I.I.; Kvetnyj, M.A.; Loginov, D.A.

1985-01-01

A new mechanism of hydraulic instability of boiling channels with convection heating which reveals in the presence of a developed ballast zone at decreased loads of a counterflan steam generator operation is considered. It is shown that for the certain combinations of thermal and technical parameters pulsation regimes caused by the ballast zone displacement over the heating surface are possible. The parameter relation at which the ballast zone position becomes unstable is obtained. The effect of the ballast zone on the statis steam generator stability is established. A mechanism of whole-circuit pulsations revealed when developing start regimes of single-pass steam generator heated with liquid sodium is explained from the positions of the instability

Topical perfluorodecalin resolves immediate whitening reactions and allows rapid effective multiple pass treatment of tattoos.

Science.gov (United States)

Reddy, Kavitha K; Brauer, Jeremy A; Anolik, Robert; Bernstein, Leonard; Brightman, Lori; Hale, Elizabeth; Karen, Julie; Weiss, Elliot; Geronemus, Roy G

2013-02-01

Laser tattoo removal using multiple passes per session, with each pass delivered after spontaneous resolution of whitening, improves tattoo fading in a 60-minute treatment time. Our objective was to evaluate the safety and efficacy of topical perfluorodecalin (PFD) in facilitating rapid effective multiple-pass tattoo removal. In a randomized, controlled study using Q-switched ruby or Nd:YAG laser, 22 previously treated tattoos were treated with 3 passes using PFD to resolve whitening after each pass ("R0 method"). In previously untreated symmetric tattoos, seven were treated over half of the tattoo with the R20 method, and the opposite half with 4 passes using PFD (R0 method); two were treated over half with a single pass and the opposite half with 4 passes using PFD (R0 method); and six treated over half with a single pass followed by PFD and the opposite half with a single pass alone. Blinded dermatologists rated tattoo fading at 1-3 months. Optical coherence tomography (OCT) imaging of whitening was performed in two tattoos. Topical PFD clinically resolved immediate whitening reactions within a mean 5 seconds (range 3-10 seconds). Tattoos treated with the R0 method demonstrated excellent fading in an average total treatment time of 5 minutes. Tattoo areas treated with the R0 method demonstrated equal fading compared to the R20 method, and improved fading compared to a single pass method. OCT imaging of whitening demonstrated epidermal and dermal hyper-reflective "bubbles" that dissipated until absent at 9-10 minutes after PFD application, and at 20 minutes without intervention. Multiple-pass tattoo removal using PFD to deliver rapid sequential passes (R0 method) appears equally effective as the R20 method, in a total treatment time averaging 5 minutes, and more effective than single pass treatment. OCT-visualized whitening-associated "bubbles," upon treatment with PFD, resolve twice as rapidly as spontaneous resolution. Copyright © 2012 Wiley

Single Pass Collider Memo: Gradient Perturbations of the SLC arc

Energy Technology Data Exchange (ETDEWEB)

Weng, W.T.; Sands, M.; /SLAC

2016-12-16

As the beam passes through the arcs, the gradient it encounters at each magnet differs from the design value. This deviation may be in part random and in part systematic. In this note we make estimates of the effects to be expected from both kinds of errors.

Single Pass Stripline Beam Position Monitor Design, Fabrication and Commissioning

Directory of Open Access Journals (Sweden)

McKinlay J.

2012-10-01

Full Text Available To monitor the position of the electron beam during transport from the Booster Synchrotron to the Storage Ring at the Australian Synchrotron, a stripline Beam Position Monitor (BPM has been designed, fabricated and installed in-house. The design was based on an existing stripline in the Booster and modified for the transfer line with a particular emphasis on ensuring the line impedance is properly matched to the detector system. The initial bench tests of a prototype stripline showed that the fabrication of the four individual striplines in the BPM was made precisely, each with a measured standing wave ratio (SWR of 1.8 at 500 MHz. Further optimization for impedance matching will be done for new stripline BPMs. The linearity and gain factor was measured with the detector system. The detector system that digitizes the signals is an Instrumentation Technologies Brilliance Single Pass [1]. The results show an error of 1 mm at an offset (from the electrical centre of 10 mm when a linear gain factor is assumed and an RMS noise of ~150 um that decreases to < 10 um with increasing signal intensity. The results were under our requirements for the transport line. The commissioning results of the stripline will also be presented showing a strong signal for an electron beam with an estimated integrated charge of ~50 nC with a position stability of 28 um (horizontal and 75 um (vertical.

Single Pass Stripline Beam Position Monitor Design, Fabrication and Commissioning

Science.gov (United States)

Tan, Y.-R. E.; Wang, D.; Van Garderen, E.; McKinlay, J.

2012-10-01

To monitor the position of the electron beam during transport from the Booster Synchrotron to the Storage Ring at the Australian Synchrotron, a stripline Beam Position Monitor (BPM) has been designed, fabricated and installed in-house. The design was based on an existing stripline in the Booster and modified for the transfer line with a particular emphasis on ensuring the line impedance is properly matched to the detector system. The initial bench tests of a prototype stripline showed that the fabrication of the four individual striplines in the BPM was made precisely, each with a measured standing wave ratio (SWR) of 1.8 at 500 MHz. Further optimization for impedance matching will be done for new stripline BPMs. The linearity and gain factor was measured with the detector system. The detector system that digitizes the signals is an Instrumentation Technologies Brilliance Single Pass [1]. The results show an error of 1 mm at an offset (from the electrical centre) of 10 mm when a linear gain factor is assumed and an RMS noise of ~150 um that decreases to < 10 um with increasing signal intensity. The results were under our requirements for the transport line. The commissioning results of the stripline will also be presented showing a strong signal for an electron beam with an estimated integrated charge of ~50 nC with a position stability of 28 um (horizontal) and 75 um (vertical).

The thermal performance of the two-pass, two-glass-cover solar air heater

Energy Technology Data Exchange (ETDEWEB)

Persad, P.; Sateunanathan, S.

1983-08-01

Analytic models are developed for the performance prediction of a two-glass-cover solar air heater operated in both the single-pass and two-pass modes. It is shown that the two-pass mode of operation is superior to the single-pass mode of operation over the range of collector inlet temperatures considered. This is seen to be mainly due to the fact that, in the two-pass mode of operation, the outer glass cover is cooled by the working fluid, thereby reducing the top losses. It is also shown that the performance in the two-pass mode of operation is independent of length, over the range of collector lengths considered, and that a critical plate spacing, dependent on the temperature level of operation of the collector, is indicated. Predicted values of performance are in good agreement with experimental results.

International Context Regarding Application of Single Failure Criterion For New Reactors

International Nuclear Information System (INIS)

Basic, I.; Vrbanic, I.

2016-01-01

The Single Failure Criterion (SFC) ensures reliable performance of safety systems in nuclear power plants in response to design basis initiating events. The SFC, basically, requires that the system must be capable of performing its task in the presence of any single failure. The capability of a system to perform its design function in the presence of a single failure could be threatened by a common cause failure such as a fire, flood, or human intervention or by any other cause with potential to induce multiple failures. When applied to plant's response to a postulated design-basis initiating event, the SFC usually represents a requirement that particular safety system performs its safety functions as designed under the conditions which can include: All failures caused by a single failure; All identifiable but non-detectable failures, including those in the non-tested components; All failures and spurious system actions that cause (or are caused by) the postulated event. The paper provides an overview of the regulatory design requirements for new reactors addressing Single Failure Criterion (SFC) in accordance to international best-practices, particularly considering the SCF relation to in-service testing, maintenance, repair, inspection and monitoring of systems, structures and components important to safety. The paper discusses the comparison of the current SFC requirements and guidelines published by the IAEA, WENRA, EUR and nuclear regulators in the United States, United Kingdom, Russia, Korea, Japan, China and Finland. Also, paper addresses the application of SFC requirements in design; considerations for testing, maintenance, repair, inspection and monitoring; allowable equipment outage times; exemptions to SFC requirements; and analysis for SFC application to two-, three- and four-train systems and applications for small and modular reactors. (author).

Bioconversion reactor

Science.gov (United States)

McCarty, Perry L.; Bachmann, Andre

1992-01-01

A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

Biphasic single-reactor process for dehydration of xylose and hydrogenation of produced furfural

NARCIS (Netherlands)

Ordomskiy, V.; Schouten, J.C.; Schaaf, van der J.; Nijhuis, T.A.

2013-01-01

The processes of xylose dehydration and the consecutive furfural hydrogenation have been combined in a single biphasic reactor. The dehydration was studied over Amberlyst-15 and the hydrogenation over a hydrophobic Ru/C catalyst. 1-Butanol, 2-methyltetrahydrofuran and cyclohexane were used as

UV-A photocatalytic treatment of Legionella pneumophila bacteria contaminated airflows through three-dimensional solid foam structured photocatalytic reactors

Energy Technology Data Exchange (ETDEWEB)

Josset, Sebastien; Hajiesmaili, Shabnam; Begin, Dominique; Edouard, David; Pham-Huu, Cuong [Laboratoire des Materiaux, Surfaces et Procedes pour la Catalyse (LMSPC), European Laboratory for Catalysis and Surface Sciences (ELCASS), CNRS, Strasbourg University, 25 rue Becquerel 67087 Strasbourg (France); Lett, Marie-Claire [Laboratoire de Genetique Moleculaire, Genomique, Microbiologie, CNRS, Strasbourg University, 28, rue Goethe 67083 Strasbourg Cedex (France); Keller, Nicolas, E-mail: nkeller@chimie.u-strasbg.fr [Laboratoire des Materiaux, Surfaces et Procedes pour la Catalyse (LMSPC), European Laboratory for Catalysis and Surface Sciences (ELCASS), CNRS, Strasbourg University, 25 rue Becquerel 67087 Strasbourg (France); Keller, Valerie [Laboratoire des Materiaux, Surfaces et Procedes pour la Catalyse (LMSPC), European Laboratory for Catalysis and Surface Sciences (ELCASS), CNRS, Strasbourg University, 25 rue Becquerel 67087 Strasbourg (France)

2010-03-15

A 3D-structured photocatalytic media was designed for allowing a tubular reactor to work in a traversing-flow mode at low pressure drops with a strong increase in the surface area-to-volume ratio inside the reactor. A protective polysiloxane coating was performed for protecting a structured polyurethane foam and anchoring the active TiO{sub 2} particles. Filled with the 3D-structured solid foam supporting TiO{sub 2} photocatalyst, the reactor could thus take advantages from the static mixer effect and from the low pressure drop resulting from the reticulated foam support. Very efficient decontamination levels towards airborne Legionella pneumophila bacteria were reached in a single-pass test mode.

Particle bed reactor nuclear rocket concept

International Nuclear Information System (INIS)

Ludewig, H.

1991-01-01

The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

A Tunable CW Orange Laser Based on a Cascaded MgO:PPLN Single-Pass Sum-Frequency Generation Module

OpenAIRE

Dismas K. Choge; Huai-Xi Chen; Bao-Lu Tian; Yi-Bin Xu; Guang-Wei Li; Wan-Guo Liang

2018-01-01

We report an all-solid-state continuous wave (CW) tunable orange laser based on cascaded single-pass sum-frequency generation with fundamental wavelengths at 1545.7 and 975.2 nm using two quasi-phase-matched (QPM) MgO-doped periodically poled lithium niobate (MgO:PPLN) crystals. Up to 10 mW of orange laser is generated in the cascaded module corresponding to a 10.4%/W nonlinear conversion efficiency. The orange output showed a temperature tuning rate of ~0.05 nm/°C, and the beam quality (M2) ...

Notes on basis band-pass circuits; Notes sur les circuits de base passe-bande

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

Resistor load amplifier stages, basic band-pass RC networks, conventional single-tuned circuits, have the same transfer function. Common properties and differences because diverse magnitude of parameters with proposed problems are exposed. Next the case of several cascaded stages (or networks) is examined when there is no reaction ones to another. (author) [French] Les etages amplificateurs a resistances, les circuits passe-bande RC elementaires, le circuit resonnant classique possedent la meme fonction de transfert. On fait ressortir les proprietes communes et les differences de comportement dues aux ordres de grandeur qu'il est possible de donner aux parametres en fonction des problemes a resoudre. On examine ensuite le cas de plusieurs etages (ou de plusieurs circuits) en cascade lorsqu'ils ne reagissent pas les uns sur les autres. (auteur)

Single-channel model for steady thermal-hydraulic analysis in nuclear reactor

International Nuclear Information System (INIS)

Zhang Xiaoying; Huang Yuanyuan

2010-01-01

This article established a single-channel model for steady analysis in the reactor and an example of thermal-hydraulic analysis was made by using this model, including the Maximum heat flux density of fuel element, enthalpy, Coolant flow, various kinds of pressure drop, enthalpy increase in average tube and thermal tube. I also got the Coolant temperature distribution and the fuel element temperature distribution and analysis of the final result. The results show that some relevant parameters which we got in this paper are well coincide with the actual operating parameters. It is also show that the single-channel model can be used to the steady thermal-hydraulic analysis. (authors)

Geometrical influences on multi-pass laser forming

International Nuclear Information System (INIS)

Edwardson, S P; Abed, E; Bartkowiak, K; Dearden, G; Watkins, K G

2006-01-01

Laser forming (LF) offers the industrial promise of controlled shaping of metallic and non-metallic components for prototyping, the correction of design shape or distortion and precision adjustment applications. The potential process advantages include precise incremental adjustment, flexibility of application and no mechanical 'spring-back' effect. To date there has been a considerable amount of work carried out on two-dimensional LF, using multi-pass straight line scan strategies to produce a reasonably controlled bend angle in a number of materials, including aerospace alloys. A key area, however, where there is a limited understanding, is the variation in bend angle per pass during multi-pass LF along a single irradiation track; in particular, the decrease in bend angle per pass after many irradiations for a given set of process parameters. Understanding this is essential if the process is to be fully controlled in a manufacturing environment. The research presented in this paper highlights the current theories as to why this occurs and proposes a further reason based on the geometrical effects of the component deformation, which in turn influences the process parameters per pass. This theory is confirmed through empirical analysis of the 2D LF process

Parametric analysis of plastic strain and force distribution in single pass metal spinning

International Nuclear Information System (INIS)

Choudhary, Shashank; Tejesh, Chiruvolu Mohan; Regalla, Srinivasa Prakash; Suresh, Kurra

2013-01-01

Metal spinning also known as spin forming is one of the sheet metal working processes by which an axis-symmetric part can be formed from a flat sheet metal blank. Parts are produced by pressing a blunt edged tool or roller on to the blank which in turn is mounted on a rotating mandrel. This paper discusses about the setting up a 3-D finite element simulation of single pass metal spinning in LS-Dyna. Four parameters were considered namely blank thickness, roller nose radius, feed ratio and mandrel speed and the variation in forces and plastic strain were analysed using the full-factorial design of experiments (DOE) method of simulation experiments. For some of these DOE runs, physical experiments on extra deep drawing (EDD) sheet metal were carried out using En31 tool on a lathe machine. Simulation results are able to predict the zone of unsafe thinning in the sheet and high forming forces that are hint to the necessity for less-expensive and semi-automated machine tools to help the household and small scale spinning workers widely prevalent in India

Parametric analysis of plastic strain and force distribution in single pass metal spinning

Science.gov (United States)

Choudhary, Shashank; Tejesh, Chiruvolu Mohan; Regalla, Srinivasa Prakash; Suresh, Kurra

2013-12-01

Metal spinning also known as spin forming is one of the sheet metal working processes by which an axis-symmetric part can be formed from a flat sheet metal blank. Parts are produced by pressing a blunt edged tool or roller on to the blank which in turn is mounted on a rotating mandrel. This paper discusses about the setting up a 3-D finite element simulation of single pass metal spinning in LS-Dyna. Four parameters were considered namely blank thickness, roller nose radius, feed ratio and mandrel speed and the variation in forces and plastic strain were analysed using the full-factorial design of experiments (DOE) method of simulation experiments. For some of these DOE runs, physical experiments on extra deep drawing (EDD) sheet metal were carried out using En31 tool on a lathe machine. Simulation results are able to predict the zone of unsafe thinning in the sheet and high forming forces that are hint to the necessity for less-expensive and semi-automated machine tools to help the household and small scale spinning workers widely prevalent in India.

Parametric analysis of plastic strain and force distribution in single pass metal spinning

Energy Technology Data Exchange (ETDEWEB)

Choudhary, Shashank, E-mail: shashankbit08@gmail.com, E-mail: mohantejesh93@gmail.com, E-mail: regalla@hyderabad.bits-pilani.ac.in, E-mail: ksuresh@hyderabad.bits-pilani.ac.in; Tejesh, Chiruvolu Mohan, E-mail: shashankbit08@gmail.com, E-mail: mohantejesh93@gmail.com, E-mail: regalla@hyderabad.bits-pilani.ac.in, E-mail: ksuresh@hyderabad.bits-pilani.ac.in; Regalla, Srinivasa Prakash, E-mail: shashankbit08@gmail.com, E-mail: mohantejesh93@gmail.com, E-mail: regalla@hyderabad.bits-pilani.ac.in, E-mail: ksuresh@hyderabad.bits-pilani.ac.in; Suresh, Kurra, E-mail: shashankbit08@gmail.com, E-mail: mohantejesh93@gmail.com, E-mail: regalla@hyderabad.bits-pilani.ac.in, E-mail: ksuresh@hyderabad.bits-pilani.ac.in [Department of Mechanical Engineering, BITS-Pilani, Hyderabad Campus, Shamirpet, Hyderabad, 500078, Andhra Pradesh (India)

2013-12-16

Metal spinning also known as spin forming is one of the sheet metal working processes by which an axis-symmetric part can be formed from a flat sheet metal blank. Parts are produced by pressing a blunt edged tool or roller on to the blank which in turn is mounted on a rotating mandrel. This paper discusses about the setting up a 3-D finite element simulation of single pass metal spinning in LS-Dyna. Four parameters were considered namely blank thickness, roller nose radius, feed ratio and mandrel speed and the variation in forces and plastic strain were analysed using the full-factorial design of experiments (DOE) method of simulation experiments. For some of these DOE runs, physical experiments on extra deep drawing (EDD) sheet metal were carried out using En31 tool on a lathe machine. Simulation results are able to predict the zone of unsafe thinning in the sheet and high forming forces that are hint to the necessity for less-expensive and semi-automated machine tools to help the household and small scale spinning workers widely prevalent in India.

Pass-transistor asynchronous sequential circuits

Science.gov (United States)

Whitaker, Sterling R.; Maki, Gary K.

1989-01-01

Design methods for asynchronous sequential pass-transistor circuits, which result in circuits that are hazard- and critical-race-free and which have added degrees of freedom for the input signals, are discussed. The design procedures are straightforward and easy to implement. Two single-transition-time state assignment methods are presented, and hardware bounds for each are established. A surprising result is that the hardware realizations for each next state variable and output variable is identical for a given flow table. Thus, a state machine with N states and M outputs can be constructed using a single layout replicated N + M times.

Correlation of histological findings of single session Er:YAG skin fractional resurfacing with various passes and energies and the possible clinical implications.

Science.gov (United States)

Trelles, Mario A; Vélez, Mariano; Mordon, Serge

2008-03-01

Ablative fractional resurfacing shows promise for skin resurfacing and tightening and also to improve treatment of epidermal and dermal pigmentary disorders. This study aimed at determining any correlation between epidermal ablation and effects on the dermis when using an Er:YAG laser in ablative fractional resurfacing mode. Ten female subjects participated in the study, mean age 52 years, Skin phototypes: 1 Fitzpatrick type II; 8 type III and 1 type IV. The degree of wrinkles (Glogau scale II or III) was similar in all cases. The laser used was the Pixel Er:YAG system (Alma Lasertrade mark, Israel) which delivers the laser beam via a hand-piece equipped with a beam splitter to divide the 2,940 nm beam into various microbeams of 850 microm in diameter in an 11 mmx11 mm treatment area. Using a constant energy of 1,400 mJ/cm(2), on a test area of 4 cmx2 cm. Two, 4, 6, and 8 passes on the preauricular area of the face were evaluated immediately after treatment. In all cases, the handpiece was kept in the same position, and rotated slightly around its perpendicular axis between passes, then moved on to the next spot. Biopsies were performed and tissue samples were routinely processed and stained with hematoxylin and eosin (H&E). No patient reported any noticeable discomfort, even at 8 passes. The histological findings revealed that, independent of the degree of the wrinkles, more laser passes produced more ablative removal of the epidermis. Residual thermal damage (RTD) with 2 laser passes was not observed but with 4 and 6 passes increased thermal effects and vacuole formation in the epidermal cells were noticed. With 8 laser passes, total epidermal removal was seen together with frank RTD-related changes in the upper part of the papillary dermis. In this study, we have demonstrated that high density fractional Er:YAG laser energy in a single session with multiple passes targeted not only the skin surface with elimination of the epidermis, but could also achieve heat

Nuclear reactor power supply

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector is interposed between the protection system and the control system. This selector prevents a parameter signal of a set of signals, which differs from the other parameters signals of the set by more than twice the allowable variation of the sensors which produce the set, from passing to the control system. The selectors include a pair of signal selection units, one unit sending selected process signals to primary control channels and the other sending selected process signals to back-up control channels. Test signals are periodically impressed by a test unit on a selected pair of a selected unit and control channels. When test signals are so impressed the selected control channel is disabled from transmitting control signals to the reactor and/or its associated components. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test

Single-phase and two-phase anaerobic digestion of fruit and vegetable waste: Comparison of start-up, reactor stability and process performance

Energy Technology Data Exchange (ETDEWEB)

Ganesh, Rangaraj [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Torrijos, Michel, E-mail: michel.torrijos@supagro.inra.fr [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Sousbie, Philippe [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Lugardon, Aurelien [Naskeo Environnment, 52 rue Paul Vaillant Couturier, F-92240 Malakoff (France); Steyer, Jean Philippe; Delgenes, Jean Philippe [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France)

2014-05-01

Highlights: • Single-phase and two-phase systems were compared for fruit and vegetable waste digestion. • Single-phase digestion produced a methane yield of 0.45 m{sup 3} CH{sub 4}/kg VS and 83% VS removal. • Substrate solubilization was high in acidification conditions at 7.0 kg VS/m{sup 3} d and pH 5.5–6.2. • Energy yield was lower by 33% for two-phase system compared to the single-phase system. • Simple and straight-forward operation favored single phase process over two-phase process. - Abstract: Single-phase and two-phase digestion of fruit and vegetable waste were studied to compare reactor start-up, reactor stability and performance (methane yield, volatile solids reduction and energy yield). The single-phase reactor (SPR) was a conventional reactor operated at a low loading rate (maximum of 3.5 kg VS/m{sup 3} d), while the two-phase system consisted of an acidification reactor (TPAR) and a methanogenic reactor (TPMR). The TPAR was inoculated with methanogenic sludge similar to the SPR, but was operated with step-wise increase in the loading rate and with total recirculation of reactor solids to convert it into acidification sludge. Before each feeding, part of the sludge from TPAR was centrifuged, the centrifuge liquid (solubilized products) was fed to the TPMR and centrifuged solids were recycled back to the reactor. Single-phase digestion produced a methane yield of 0.45 m{sup 3} CH{sub 4}/kg VS fed and VS removal of 83%. The TPAR shifted to acidification mode at an OLR of 10.0 kg VS/m{sup 3} d and then achieved stable performance at 7.0 kg VS/m{sup 3} d and pH 5.5–6.2, with very high substrate solubilization rate and a methane yield of 0.30 m{sup 3} CH{sub 4}/kg COD fed. The two-phase process was capable of high VS reduction, but material and energy balance showed that the single-phase process was superior in terms of volumetric methane production and energy yield by 33%. The lower energy yield of the two-phase system was due to the loss of

Single-phase and two-phase anaerobic digestion of fruit and vegetable waste: Comparison of start-up, reactor stability and process performance

International Nuclear Information System (INIS)

Ganesh, Rangaraj; Torrijos, Michel; Sousbie, Philippe; Lugardon, Aurelien; Steyer, Jean Philippe; Delgenes, Jean Philippe

2014-01-01

Highlights: • Single-phase and two-phase systems were compared for fruit and vegetable waste digestion. • Single-phase digestion produced a methane yield of 0.45 m 3 CH 4 /kg VS and 83% VS removal. • Substrate solubilization was high in acidification conditions at 7.0 kg VS/m 3 d and pH 5.5–6.2. • Energy yield was lower by 33% for two-phase system compared to the single-phase system. • Simple and straight-forward operation favored single phase process over two-phase process. - Abstract: Single-phase and two-phase digestion of fruit and vegetable waste were studied to compare reactor start-up, reactor stability and performance (methane yield, volatile solids reduction and energy yield). The single-phase reactor (SPR) was a conventional reactor operated at a low loading rate (maximum of 3.5 kg VS/m 3 d), while the two-phase system consisted of an acidification reactor (TPAR) and a methanogenic reactor (TPMR). The TPAR was inoculated with methanogenic sludge similar to the SPR, but was operated with step-wise increase in the loading rate and with total recirculation of reactor solids to convert it into acidification sludge. Before each feeding, part of the sludge from TPAR was centrifuged, the centrifuge liquid (solubilized products) was fed to the TPMR and centrifuged solids were recycled back to the reactor. Single-phase digestion produced a methane yield of 0.45 m 3 CH 4 /kg VS fed and VS removal of 83%. The TPAR shifted to acidification mode at an OLR of 10.0 kg VS/m 3 d and then achieved stable performance at 7.0 kg VS/m 3 d and pH 5.5–6.2, with very high substrate solubilization rate and a methane yield of 0.30 m 3 CH 4 /kg COD fed. The two-phase process was capable of high VS reduction, but material and energy balance showed that the single-phase process was superior in terms of volumetric methane production and energy yield by 33%. The lower energy yield of the two-phase system was due to the loss of energy during hydrolysis in the TPAR and the

Design of a supercritical water-cooled reactor. Pressure vessel and internals

Energy Technology Data Exchange (ETDEWEB)

Fischer, Kai

2008-08-15

The High Performance Light Water Reactor (HPLWR) is a light water reactor with supercritical steam conditions which has been investigated within the 5th Framework Program of the European Commission. Due to the supercritical pressure of 25 MPa, water, used as moderator and as coolant, flows as a single phase through the core and can be directly fed to the turbine. Using the technology of coal fired power plants with supercritical steam conditions, the heat-up in the core is done in several steps to achieve the targeted high steam outlet temperature of 500.C without exceeding available cladding material limits. Based on a first design of a fuel assembly cluster for a HPLWR with a single pass core, the surrounding internals and the reactor pressure vessel (RPV) are dimensioned for the first time, following the safety standards of the nuclear safety standards commission in Germany. Furthermore, this design is extended to the incorporation of core arrangements with two and three passes. The design of the internals and the RPV are verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Additionally, a passive safety component for the feedwater inlet of the RPV of the HPLWR is designed. Its purpose is the reduction of the mass flow rate in case of a LOCA for a feedwater line break until further steps are executed. Starting with a simple vortex diode, several steps are executed to enhance the performance of the diode and adapt it to this application. Then, this first design is further optimized using combined 1D and 3D flow analyses. Parametric studies determine the performance and characteristic for changing mass flow rates for this backflow limiter. (orig.)

A Single-Granule-Level Approach Reveals Ecological Heterogeneity in an Upflow Anaerobic Sludge Blanket Reactor.

Directory of Open Access Journals (Sweden)

Kyohei Kuroda

Full Text Available Upflow anaerobic sludge blanket (UASB reactor has served as an effective process to treat industrial wastewater such as purified terephthalic acid (PTA wastewater. For optimal UASB performance, balanced ecological interactions between syntrophs, methanogens, and fermenters are critical. However, much of the interactions remain unclear because UASB have been studied at a "macro"-level perspective of the reactor ecosystem. In reality, such reactors are composed of a suite of granules, each forming individual micro-ecosystems treating wastewater. Thus, typical approaches may be oversimplifying the complexity of the microbial ecology and granular development. To identify critical microbial interactions at both macro- and micro- level ecosystem ecology, we perform community and network analyses on 300 PTA-degrading granules from a lab-scale UASB reactor and two full-scale reactors. Based on MiSeq-based 16S rRNA gene sequencing of individual granules, different granule-types co-exist in both full-scale reactors regardless of granule size and reactor sampling depth, suggesting that distinct microbial interactions occur in different granules throughout the reactor. In addition, we identify novel networks of syntrophic metabolic interactions in different granules, perhaps caused by distinct thermodynamic conditions. Moreover, unseen methanogenic relationships (e.g. "Candidatus Aminicenantes" and Methanosaeta are observed in UASB reactors. In total, we discover unexpected microbial interactions in granular micro-ecosystems supporting UASB ecology and treatment through a unique single-granule level approach.

A Single-Granule-Level Approach Reveals Ecological Heterogeneity in an Upflow Anaerobic Sludge Blanket Reactor

Science.gov (United States)

Mei, Ran; Narihiro, Takashi; Bocher, Benjamin T. W.; Yamaguchi, Takashi; Liu, Wen-Tso

2016-01-01

Upflow anaerobic sludge blanket (UASB) reactor has served as an effective process to treat industrial wastewater such as purified terephthalic acid (PTA) wastewater. For optimal UASB performance, balanced ecological interactions between syntrophs, methanogens, and fermenters are critical. However, much of the interactions remain unclear because UASB have been studied at a “macro”-level perspective of the reactor ecosystem. In reality, such reactors are composed of a suite of granules, each forming individual micro-ecosystems treating wastewater. Thus, typical approaches may be oversimplifying the complexity of the microbial ecology and granular development. To identify critical microbial interactions at both macro- and micro- level ecosystem ecology, we perform community and network analyses on 300 PTA–degrading granules from a lab-scale UASB reactor and two full-scale reactors. Based on MiSeq-based 16S rRNA gene sequencing of individual granules, different granule-types co-exist in both full-scale reactors regardless of granule size and reactor sampling depth, suggesting that distinct microbial interactions occur in different granules throughout the reactor. In addition, we identify novel networks of syntrophic metabolic interactions in different granules, perhaps caused by distinct thermodynamic conditions. Moreover, unseen methanogenic relationships (e.g. “Candidatus Aminicenantes” and Methanosaeta) are observed in UASB reactors. In total, we discover unexpected microbial interactions in granular micro-ecosystems supporting UASB ecology and treatment through a unique single-granule level approach. PMID:27936088

Economic Optimizing Control for Single-Cell Protein Production in a U-Loop Reactor

DEFF Research Database (Denmark)

Drejer, André; Ritschel, Tobias Kasper Skovborg; Jørgensen, Sten Bay

2017-01-01

The production of single-cell protein (SCP) in a U-loop reactor by a methanotroph is a cost efficient sustainable alternative to protein from fish meal obtained by over-fishing the oceans. SCP serves as animal feed. In this paper, we present a mathematical model that describes the dynamics of SCP...

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

Selective purge for hydrogenation reactor recycle loop

Science.gov (United States)

Baker, Richard W.; Lokhandwala, Kaaeid A.

2001-01-01

Processes and apparatus for providing improved contaminant removal and hydrogen recovery in hydrogenation reactors, particularly in refineries and petrochemical plants. The improved contaminant removal is achieved by selective purging, by passing gases in the hydrogenation reactor recycle loop or purge stream across membranes selective in favor of the contaminant over hydrogen.

Intestinal permeability of forskolin by in situ single pass perfusion in rats.

Science.gov (United States)

Liu, Zhen-Jun; Jiang, Dong-bo; Tian, Lu-Lu; Yin, Jia-Jun; Huang, Jian-Ming; Weng, Wei-Yu

2012-05-01

The intestinal permeability of forskolin was investigated using a single pass intestinal perfusion (SPIP) technique in rats. SPIP was performed in different intestinal segments (duodenum, jejunum, ileum, and colon) with three concentrations of forskolin (11.90, 29.75, and 59.90 µg/mL). The investigations of adsorption and stability were performed to ensure that the disappearance of forskolin from the perfusate was due to intestinal absorption. The results of the SPIP study indicated that forskolin could be absorbed in all segments of the intestine. The effective permeability (P (eff)) of forskolin was in the range of drugs with high intestinal permeability. The P (eff) was highest in the duodenum as compared to other intestinal segments. The decreases of P (eff) in the duodenum and ileum at the highest forskolin concentration suggested a saturable transport process. The addition of verapamil, a P-glycoprotein inhibitor, significantly enhanced the permeability of forskolin across the rat jejunum. The absorbed fraction of dissolved forskolin after oral administration in humans was estimated to be 100 % calculated from rat P (eff). In conclusion, dissolved forskolin can be absorbed readily in the intestine. The low aqueous solubility of forskolin might be a crucial factor for its poor oral bioavailability. © Georg Thieme Verlag KG Stuttgart · New York.

Single-crystal filters for attenuating epithermal neutrons and gamma rays in reactor beams

DEFF Research Database (Denmark)

Rustad, B.M.; Als-Nielsen, Jens Aage; Bahnsen, A.

1965-01-01

Cross section of representative samples of bismuth and quartz were measured at room and liquid nitrogen temperatures over neutron energy range of 0.0007 to 2.0 ev to obtain data for design of single-crystal 32-cm bismuth filters for attenuating fast neutrons and γ-rays in reactor beams; filters may...

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Full-Scale Continuous Mini-Reactor Setup for Heterogeneous Grignard Alkylation of a Pharmaceutical Intermediate

DEFF Research Database (Denmark)

Pedersen, Michael Jønch; Holm, Thomas; Rahbek, Jesper P.

2013-01-01

A reactor setup consisting of two reactors in series has been implemented for a full-scale, heterogeneous Grignard alkylation. Solutions pass from a small filter reactor into a static mixer reactor with multiple side entries, thus combining continuous stirred tank reactor (CSTR) and plug flow...

Prediction of TRISO coated particle performances for a one-pass deep burn

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

2008-02-15

In the present studies, TRISO coated particle performances have been investigated for incinerating plutonium and minor actinides by the Gas Turbine-Modular Helium Reactor, whose fresh fuel is fabricated after the uranium extraction (UREX) process applied to Light Water Reactors irradiated fuel. The analyses divide into two parts: in the first part, the latest design of the reactor core proposed by General Atomics, which takes advantage of four fuel rings, has been modeled in deep details by the Monte Carlo MCNP code and a burnup process has been simulated by the MCB code. In the second part, the TRISO coated particle performances have been investigated by the PANAMA code with the goal of verifying the design constraints proposed by General Atomics. During burnup, the refueling and shuffling schedule followed the one-pass deep burn concept, where the fuel is utilized, since fabrication for the Gas Turbine-Modular Helium Reactor, without any reprocessing until the final disposal into the geological repository. During the reactor operation, the fast fluence on all TRISO particles layers has been evaluated and the production of the key fission products monitored. During an hypothetical reactor accident scenario, the TRISO particle failure fraction has been estimated.

Prediction of TRISO coated particle performances for a one-pass deep burn

International Nuclear Information System (INIS)

Talamo, Alberto

2008-01-01

In the present studies, TRISO coated particle performances have been investigated for incinerating plutonium and minor actinides by the Gas Turbine-Modular Helium Reactor, whose fresh fuel is fabricated after the uranium extraction (UREX) process applied to Light Water Reactors irradiated fuel. The analyses divide into two parts: in the first part, the latest design of the reactor core proposed by General Atomics, which takes advantage of four fuel rings, has been modeled in deep details by the Monte Carlo MCNP code and a burnup process has been simulated by the MCB code. In the second part, the TRISO coated particle performances have been investigated by the PANAMA code with the goal of verifying the design constraints proposed by General Atomics. During burnup, the refueling and shuffling schedule followed the one-pass deep burn concept, where the fuel is utilized, since fabrication for the Gas Turbine-Modular Helium Reactor, without any reprocessing until the final disposal into the geological repository. During the reactor operation, the fast fluence on all TRISO particles layers has been evaluated and the production of the key fission products monitored. During an hypothetical reactor accident scenario, the TRISO particle failure fraction has been estimated

Single-pass high-gain tapered free-electron laser with transverse diffraction in the postsaturation regime

Directory of Open Access Journals (Sweden)

Cheng-Ying Tsai

2018-06-01

Full Text Available It has been well known that the resonant interaction of an ultrarelativistic electron beam and the radiation field in the single-pass high-gain free electron laser (FEL amplifier leads to the optical gain guiding. The transverse Laplacian term of the slowly varying wave equation in the linear regime can be approximated as a constant detuning parameter, i.e., |∇_{⊥}^{2}|∼k_{R}/z_{R} where k_{R} is the resonant wave number and z_{R} is the Rayleigh range of the laser. In the post-saturation regime, the radiation power begins to oscillate about an equilibrium for the untapered case while continues to grow by undulator tapering. Moreover, in this regime the gain guiding decreases and the simple constant detune is no longer valid. In this paper we study the single-pass high-gain FEL performance in the post-saturation regime with inclusion of diffraction effect and undulator tapering. Our analysis relies upon two constants of motion, one from the energy conservation and the other from the adiabatic invariant of the action variable. By constructing a two-dimensional axisymmetric wave equation and the coupled one-dimensional electron dynamical equations, the performance of a tapered FEL in the postsaturation regime can be analyzed, including the fundamental mode profile, the power efficiency and the scaled energy spread. We begin the analytical investigation with two different axisymmetric electron beam profiles, the uniform and bounded parabolic ones. It is found that the tapered FEL power efficiency can be smaller but close to the taper ratio provided the resonant phase remains constant and the beam-wave is properly matched. Such a tapered efficiency is nearly independent of transverse electron beam size before significant electron detrapping occurs. This is essentially different from the untapered case, where the power extraction efficiency is around the essential FEL gain bandwidth (or ρ, the Pierce or FEL parameter and depends on the beam

Message Passing Framework for Globally Interconnected Clusters

International Nuclear Information System (INIS)

Hafeez, M; Riaz, N; Asghar, S; Malik, U A; Rehman, A

2011-01-01

In prevailing technology trends it is apparent that the network requirements and technologies will advance in future. Therefore the need of High Performance Computing (HPC) based implementation for interconnecting clusters is comprehensible for scalability of clusters. Grid computing provides global infrastructure of interconnecting clusters consisting of dispersed computing resources over Internet. On the other hand the leading model for HPC programming is Message Passing Interface (MPI). As compared to Grid computing, MPI is better suited for solving most of the complex computational problems. MPI itself is restricted to a single cluster. It does not support message passing over the internet to use the computing resources of different clusters in an optimal way. We propose a model that provides message passing capabilities between parallel applications over the internet. The proposed model is based on Architecture for Java Universal Message Passing (A-JUMP) framework and Enterprise Service Bus (ESB) named as High Performance Computing Bus. The HPC Bus is built using ActiveMQ. HPC Bus is responsible for communication and message passing in an asynchronous manner. Asynchronous mode of communication offers an assurance for message delivery as well as a fault tolerance mechanism for message passing. The idea presented in this paper effectively utilizes wide-area intercluster networks. It also provides scheduling, dynamic resource discovery and allocation, and sub-clustering of resources for different jobs. Performance analysis and comparison study of the proposed framework with P2P-MPI are also presented in this paper.

Measuring device for the coolant flowrate in a reactor core

International Nuclear Information System (INIS)

Sawa, Toshihiko.

1983-01-01

Purpose: To improve the operation performance by enabling direct and accurate measurement for the reactor core recycling flowrate. Constitution: A control rod guide is disposed to the upper end of a control rod drive mechanism housing passing through the bottom of a reactor pressure vessel and it is inserted into the through hole of a reactor core support plate. A water flow passage is formed through the reactor core support plate for the flowrate measurement of coolants recycled within the reactor core. The static pressure difference between the upper and the lower sides of the reactor core support plate is measured by a pressure difference detector of a pressure difference measuring mechanism, and an output signal from the pressure different detector is inputted to a calculation means, in which the amount of the coolants passing through the water flow passage is calculated based on the output signal corresponding to the pressure difference. Then, the total recycling flowrate in the reactor core is determined in the calculation means based on the relation between the measured flowrate and a predetermined total reactor core recycling flowrate. (Horiuchi, T.)

Single-piece maintenance procedures for the TITAN reversed-field pinch reactor

International Nuclear Information System (INIS)

Grotz, S.P.; Creedon, R.L.; Cooke, P.I.H.; Duggan, W.P.; Krakowski, R.A.; Najmabadi, F.; Wong, C.P.C.

1987-01-01

The TITAN reactor is a compact (major radius of 3.9 m and minor plasma radius of 0.6 m), high neutron wall loading (--18MW/m 2 ) fusion energy system based on the reversed-field pinch (RFP) concept. The TITAN-I fusion power core (FPC) is a lithium, self-cooled design with vanadium alloy (V-3Ti-1Si) structural material. The compact design of the TITAN fusion power core (FPC) reduces the system to a few small and relatively low mass components, making toroidal segmentation of the FPC unnecessary. A single-piece maintenance procedure in which the replaceable first wall and blanket is removed as a single unit is, therefore, possible. The TITAN FPC design provides for top access to the reactor with vertical lifts used to remove the components. The number of remote handling procedures is few and the movements are uncomplicated. The annual torus replacement requires that the reusable ohmic-heating coil set and hot-shield assembly be removed and temporarily stored in a hot cell. The used first wall and blanket assembly is drained and disconnected from the coolant supply system, then lifted to a processing room where it is cooled and prepared for Class-C waste burial. The new, pre-tested first wall and blanket assembly is then lowered into position and the removal procedure is reversed to complete the replacement process

Process Stability Identification Through Dynamic Study of Single-bed Ammonia Reactor with Feed-Effluent Heat Exchanger (FEHE

Directory of Open Access Journals (Sweden)

Adhi Tri Partono

2018-01-01

Full Text Available In ammonia reactor system, a feed-effluent heat exchanger (FEHE is typically installed to utilize reaction-generated heat to heat the reactor’s feed. Utilizing energy from exothermic reaction to the incoming feed stream is often called “autothermic operation”. Despite the advantage of FEHE, there is a risk of utilizing FEHE in a reactor system such as instability of process temperature or known as hysteresis. Hysteresis phenomena in chemical process could cause operational problems, for example it could damage the integrity of the equipment’s material. This paper aims to evaluate the dynamic behavior of a single-bed ammonia reactor with FEHE, particularly to propose a way to prevent instability within the system. The dynamic simulation of the single-bed ammonia reactor with FEHE was performed with Aspen HYSYS v8.8. The result of the simulation result shows that hysteresis phenomenon in the ammonia reactor system occurs when the feed’s temperature is below a certain value. If the feed temperature reaches that value, the temperature of the reactor’s outlet oscillates. One of the solution to keep the feed temperature above that critical value is by installing a trim heater within the system. Based on the simulation, trim heater installation within the system is able to prevent hysteresis in the system evaluated.

European developments in single phase turbulence for innovative reactors

Energy Technology Data Exchange (ETDEWEB)

Roelofs, F., E-mail: roelofs@nrg.eu [NRG, Petten (Netherlands); Rohde, M. [DUT, Delft (Netherlands); and others

2011-07-01

Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). They result in specific behavior of flow and heat transfer, which requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulics topics are the motivation for the THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which is sponsored by the European Commission from 2010 to 2014. This paper describes the ongoing developments in an important part of this project devoted to single phase turbulence issues. To this respect, the two main issues have been identified: Non-unity Prandtl number turbulence. In case of liquid metals, molten salts or supercritical fluids, the commonly applied constant turbulent Prandtl number concept is not applicable and robust engineering turbulence models are needed. This paper will report on the progress achieved with respect to the development and validation of turbulence models available in commonly used engineering tools. The paper also reports about the supporting experiments and direct numerical simulations; and, Temperature fluctuations possibly leading to thermal fatigue in innovative reactors. The status is described of a fundamental experiment dealing with the mixing of different density gases in a rectangular channel, an experiment in a more complex geometry of a small mixing plenum using a supercritical fluid, and direct numerical simulations of conjugate heat transfer on temperature fluctuations in liquid metal. (author)

European developments in single phase turbulence for innovative reactors

International Nuclear Information System (INIS)

Roelofs, F.; Rohde, M.

2011-01-01

Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). They result in specific behavior of flow and heat transfer, which requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulics topics are the motivation for the THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which is sponsored by the European Commission from 2010 to 2014. This paper describes the ongoing developments in an important part of this project devoted to single phase turbulence issues. To this respect, the two main issues have been identified: Non-unity Prandtl number turbulence. In case of liquid metals, molten salts or supercritical fluids, the commonly applied constant turbulent Prandtl number concept is not applicable and robust engineering turbulence models are needed. This paper will report on the progress achieved with respect to the development and validation of turbulence models available in commonly used engineering tools. The paper also reports about the supporting experiments and direct numerical simulations; and, Temperature fluctuations possibly leading to thermal fatigue in innovative reactors. The status is described of a fundamental experiment dealing with the mixing of different density gases in a rectangular channel, an experiment in a more complex geometry of a small mixing plenum using a supercritical fluid, and direct numerical simulations of conjugate heat transfer on temperature fluctuations in liquid metal. (author)

Control of PWR reactor energy supplied to a stream turbine

International Nuclear Information System (INIS)

Petetrot, J.F.; Parent, Pierre.

1981-01-01

This patent presents a process for regulating the power provided by a pressurized water nuclear reactor to a steam turbine, by moving the control rods absorbing the neutrons in the reactor core and by diverting a fraction of the steam produced by the reactor, outside the turbine circuit, by opening by-pass valves [fr

The role of research reactor and its future

International Nuclear Information System (INIS)

Nakagome, Yoshihiro

2005-01-01

About a half century passed since the start of operation of research reactors. Many research reactors were stopped their operation or decommissioned. With the practical use of nuclear energy, the meaning of research reactor has been buried in oblivion in the developed countries. Furthermore, under the nuclear weapons nonproliferation policy, the use of high enriched uranium fuel in research reactors is obliged to change to the use of low enriched uranium fuel. In such severe situation, this paper refers to the role of the research reactor once more through the operation experience of university-owned research reactor KUR (Kyoto University Reactor, Japan) and describes that research reactor is indispensable for the preparation to the second coming nuclear age. (author)

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

Pebble-bed reactor

International Nuclear Information System (INIS)

Lohnert, G.; Mueller-Frank, U.; Heil, J.

1976-01-01

A pebble-bed nuclear reactor of large power rating comprises a container having a funnel-shaped bottom forming a pebble run-out having a centrally positioned outlet. A bed of downwardly-flowing substantially spherical nuclear fuel pebbles is positioned in the container and forms a reactive nuclear core maintained by feeding unused pebbles to the bed's top surface while used or burned-out pebbles run out and discharge through the outlet. A substantially conical body with its apex pointing upwardly and its periphery spaced from the periphery of the container spreads the bottom of the bed outwardly to provide an annular flow down the funnel-shaped bottom forming the runout, to the discharge outlet. This provides a largely constant downward velocity of the spheres throughout the diameter of the bed throughout a substantial portion of the down travel, so that all spheres reach about the same burned-out condition when they leave the core, after a single pass through the core area

Air conditioning device for reactor buildings

International Nuclear Information System (INIS)

Kikuchi, Shiro.

1982-01-01

Purpose: To decrease the opening areas of pipe lines for an air conditioning device at the portions passing through the shielding walls of a reactor building for a FBR type reactor, as well as reduce the size of the building. Constitution: Airs in the building for containing reactor are liquefied in an air liquefying mechanism. The liquefied airs are sent by way of pipe lines to each of evaporators, wherein each of the chambers are cooled because of latent heat of evaporation and evaporated airs are released to each of the chambers. The airs released to each of the chambers are collected into an exhaust chamber and sent by way of a duct to the air liquefying mechanism and liquefied again. Since the volume of the liquefied airs may be smaller than the amount conventionally required for usual cooled airs, the pipe lines passing through the shielding walls of the building can be of smaller diameter. This can decrease the opening areas of the pipe lines at the portions passing through the walls of the shieldings and, since the opening areas are smaller, the structure of the radiation shieldings can be simplified in these portions. Further, since the space of the pipe lines in the building is reduced extremely, the size of the building can be reduced. (Moriyama, K.)

Gas-phase optical fiber photocatalytic reactors for indoor air application: a preliminary study on performance indicators

Science.gov (United States)

Palmiste, Ü.; Voll, H.

2017-10-01

The development of advanced air cleaning technologies aims to reduce building energy consumption by reduction of outdoor air flow rates while keeping the indoor air quality at an acceptable level by air cleaning. Photocatalytic oxidation is an emerging technology for gas-phase air cleaning that can be applied in a standalone unit or a subsystem of a building mechanical ventilation system. Quantitative information on photocatalytic reactor performance is required to evaluate the technical and economic viability of the advanced air cleaning by PCO technology as an energy conservation measure in a building air conditioning system. Photocatalytic reactors applying optical fibers as light guide or photocatalyst coating support have been reported as an approach to address the current light utilization problems and thus, improve the overall efficiency. The aim of the paper is to present a preliminary evaluation on continuous flow optical fiber photocatalytic reactors based on performance indicators commonly applied for air cleaners. Based on experimental data, monolith-type optical fiber reactor performance surpasses annular-type optical fiber reactors in single-pass removal efficiency, clean air delivery rate and operating cost efficiency.

Comparison of reactivity in a flow reactor and a single cylinder engine

Energy Technology Data Exchange (ETDEWEB)

Natelson, Robert H.; Johnson, Rodney O.; Kurman, Matthew S.; Cernansky, Nicholas P.; Miller, David L. [Department of Mechanical Engineering and Mechanics, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104-2875 (United States)

2010-10-15

The relative reactivity of 2:1:1 and 1:1:1 mixtures of n-decane:n-butylcyclohexane:n-butylbenzene and an average sample of JP-8 were evaluated in a single cylinder engine and compared to results obtained in a pressurized flow reactor. At compression ratios of 14:1, 15:1, and 16:1, inlet temperature of 500 K, inlet pressure of 0.1 MPa, equivalence ratio of 0.23, and engine speed of 800 RPM, the autoignition delay times were, from shortest to longest, the 2:1:1, followed by the 1:1:1, and then the JP-8. This order corresponded with recent results in a pressurized flow reactor, where the preignition oxidation chemistry was monitored at temperatures of 600-800 K, 0.8 MPa pressure, and an equivalence ratio of 0.30, and where the preignition reactivity from highest to lowest was the 2:1:1, followed by the 1:1:1, and the JP-8. This shows that the relative reactivity at low temperatures in the flow reactor tracks the autoignition tendencies in the engine for these particular fuels. (author) the computed experimental error. (author)

Monitoring device for the reactor pipelines

International Nuclear Information System (INIS)

Fukumoto, Akira.

1983-01-01

Purpose: To enable rapid and accurate operator's monitoring for the state of pipelines in a BWR type reactor. Constitution: Specific symbols are attached respectively to a fluid supply source constituting the pipelines of a nuclear reactor facility, a plurality of fluid passing points and equipments to be supplied with the fluid, and a symmetrical matrix comprising these symbols in rows and columns is constituted. Then, a matrix is prepared based on detection signals for the states of the liquid supply source, equipments to be supplied with fluid and pipeline equipments by rendering the matrix elements between the signals expressing the state capable of passing the fluid as 1 and the matrix elements between the signals expressing the state incapable of passing the fluid as 0 . The matrix thus prepared in a signal procession circuit and a matrix in a memory circuit previously storing the matrix expressing the normal state of the pipelines are compared to judge the state of the pipelines in a short time and with no misjudging. (Moriyama, K.)

Incorporating single detector failure into the ROP detector layout optimization for CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Kastanya, Doddy, E-mail: Doddy.Kastanya@snclavalin.com

2015-12-15

Highlights: • ROP TSP value needs to be adjusted when any detector in the system fails. • Single detector failure criterion has been incorporated into the detector layout optimization as a constraint. • Results show that the optimized detector layout is more robust with respect to its vulnerability to a single detector failure. • An early rejection scheme has been introduced to speed-up the optimization process. - Abstract: In CANDU{sup ®} reactors, the regional overpower protection (ROP) systems are designed to protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. In the CANDU{sup ®} 600 MW (CANDU 6) design, there are two ROP systems in the core, each of which is connected to a fast-acting shutdown system. Each ROP system consists of a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal flux detector assemblies. The placement of these ROP detectors is a challenging discrete optimization problem. In the past few years, two algorithms, DETPLASA and ADORE, have been developed to optimize the detector layout for the ROP systems in CANDU reactors. These algorithms utilize the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The objective of the optimization process is typically either to maximize the TSP value for a given number of detectors in the system or to minimize the number of detectors in the system to obtain a target TSP value. One measure to determine the robustness of the optimized detector layout is to evaluate the maximum decrease (penalty) in TSP value when any single detector in the system fails. The smaller the penalty, the more robust the design is. Therefore, in order to ensure that the optimized detector layout is robust, the single detector failure (SDF) criterion has been incorporated as an additional constraint into the ADORE algorithm. Results from this study indicate that there

Sources of variability for the single-comparator method in a heavy-water reactor

International Nuclear Information System (INIS)

Damsgaard, E.; Heydorn, K.

1978-11-01

The well thermalized flux in the heavy-water-moderated DR 3 reactor at Risoe prompted us to investigate to what extent a single comparator could be used for multi-element determination instead of multiple comparators. The reliability of the single-comparator method is limited by the thermal-to-epithermal ratio, and experiments were designed to determine the variations in this ratio throughout a reactor operating period (4 weeks including a shut-down period of 4-5 days). The bi-isotopic method using zirconium as monitor was chosen, because 94 Zr and 96 Zr exhibit a large difference in their Isub(o)/Σsub(th) values, and would permit determination of the flux ratio with a precision sufficient to determine variations. One of the irradiation facilities comprises a rotating magazine with 3 channels, each of which can hold five aluminium cans. In this rig, five cans, each holding a polyvial with 1 ml of aqueous zirconium solution were irradiated simultaneously in one channel. Irradiations were carried out in the first and the third week of 4 periods. In another facility consisting of a pneumatic tube system, two samples were simultaneously irradiated on top of each other in a polyethylene rabbit. Experiments were carried out once a week for 4 periods. All samples were counted on a Ge(Li)-detector for 95 Zr, 97 sup(m)Nb and 97 Nb. The thermal-to-epithermal flux ratio was calculated from the induced activity, the nuclear data for the two zirconium isotopes and the detector efficiency. By analysis of variance the total variation of the flux ratio was separated into a random variation between reactor periods, and systematic differences between the positions, as well as the weeks in the operating period. If the variations are in statistical control, the error resulting from use of the single-comparator method in multi-element determination can be estimated for any combination of irradiation position and day in the operating period. With the measure flux ratio variations in DR

Damage of reactor buildings occurred at the Fukushima Daiichi accident. Focusing on sequence leading to hydrogen explosions

International Nuclear Information System (INIS)

Naito, Masanori

2011-01-01

Fukushima Daiichi accident discharged enormous radioactive materials confined inside into the environment due to hydrogen explosions occurred at reactor buildings and forced many people to live the refugee life. This article described overview of Great East Japan Earthquake, specifications of Fukushima Daiichi nuclear power plants, sequence of plant status after earthquake occurrence and computerized simulation of plant behavior of Unit 1 leading to core melt and hydrogen explosion. Simulation results with estimated and assumed conditions showed water level decreased to bottom of reactor core after 4 hrs and 15 minutes passed, core melt started after 6 hrs and 49 minutes passed, failure of core support plate after 7 hrs and 18 minutes passed and through failure of penetration at bottom of pressure vessel after 7 hrs and 25 minutes passed. Hydrogen concentration at operating floor of reactor building of Unit 1 would be 15% accumulated and the pressure would amount to about 5 bars after hydrogen explosion if reactor building did not rupture with leak-tight structure. Since reactor building was not pressure-proof structure, walls of operating floor would rupture before 5 bars attained. (T. Tanaka)

Design features of BREST reactors. Experimental work to advance the concept of BREST reactors. Results and plans

International Nuclear Information System (INIS)

Filin, A.I.; Orlov, V.V.; Leonov, V.N.; Sila-Novitskij, A.G.; Smirnov, V.S.; Tsikunov, V.S.

2001-01-01

Principle designs of 300 MW(th) and 1200 MW(th) lead-cooled fast reactors are presented. Reactors of various output are shown to be built using the same principles. In conjunction with increased output and to implement inherent safety concept in BREST-1200 reactor design a number of new solutions, which may be used in BREST-300 concept too, has been taken including: pool-type reactor design not requiring metal vessel, hence, not limiting reactor power; new handling system allowing to reduce central hall and building dimensions as a whole; emergency cooling system using Field pipes, immersed directly in lead, which may be used to cool down reactor under normal conditions; by-pass line incorporated in coolant loop allowing to refuse the actively actuating valve initiated in pumps shut down. (author)

The Single Pass RF Driver: Final beam compression

Energy Technology Data Exchange (ETDEWEB)

Burke, Robert, E-mail: rjburke@fusionpowercorporation.com

2014-01-01

The Single Pass RF Driver (SPRFD) compacts the beam from the linac without storage rings by manipulations that take advantage of the multiplicity of isotopes (16), the preserved µbunch structure, and increased total linac current. Magnetic switches on a first set of delay lines rearrange the internal structure of the various isotopic beams. A second set of delay lines sets the relative timing of the 16 isotopic beam sections so they will telescope at the pellet, in one of multiple fusion chambers, e.g. 10. Shortening each isotopic beam section uses preservation of the µbunch structure up to the final ∼2 km drift before final focus. Just before the final drift, differential acceleration of the µbunches in each isotopic beam section (128 total) launches an axial collapse, referred to as the “Slick”. The µbunches interpenetrate as their centers of mass move toward each other and individual µbunches lengthen due to their momentum spread. In longitudinal phase space they slide over one another as they lengthen in time and slim down in instantaneous energy spread. The permissible amount of µbunch lengthening is set by the design pulse shape at the pellet, which varies for different groups of isotopes. In narrow bands of ranges according to the role for each isotope group in the pellet, the ranges extend from 1 to 10 g/cm{sup 2} to drive the cylinder barrel and thin hemispherical end caps, to heat the ∼0.5 g/cm{sup 2}ρR fast ignition zone, and to improve the quasi-sphericity of the compression of the fast ignition zones at the pellet's ends. Because the µbunch–µbunch momentum differences are correlated, time-ramped beamline transport elements close after the differential accelerator are used to correct the associated shifts of focal point. Beam neutralization is needed after the differential acceleration until adjacent bunches begin to overlap. Concurrent collapse of each isotope and telescoping of the 16 isotopes cause the current in each beamline

The Single Pass RF Driver: Final beam compression

International Nuclear Information System (INIS)

Burke, Robert

2014-01-01

The Single Pass RF Driver (SPRFD) compacts the beam from the linac without storage rings by manipulations that take advantage of the multiplicity of isotopes (16), the preserved µbunch structure, and increased total linac current. Magnetic switches on a first set of delay lines rearrange the internal structure of the various isotopic beams. A second set of delay lines sets the relative timing of the 16 isotopic beam sections so they will telescope at the pellet, in one of multiple fusion chambers, e.g. 10. Shortening each isotopic beam section uses preservation of the µbunch structure up to the final ∼2 km drift before final focus. Just before the final drift, differential acceleration of the µbunches in each isotopic beam section (128 total) launches an axial collapse, referred to as the “Slick”. The µbunches interpenetrate as their centers of mass move toward each other and individual µbunches lengthen due to their momentum spread. In longitudinal phase space they slide over one another as they lengthen in time and slim down in instantaneous energy spread. The permissible amount of µbunch lengthening is set by the design pulse shape at the pellet, which varies for different groups of isotopes. In narrow bands of ranges according to the role for each isotope group in the pellet, the ranges extend from 1 to 10 g/cm 2 to drive the cylinder barrel and thin hemispherical end caps, to heat the ∼0.5 g/cm 2 ρR fast ignition zone, and to improve the quasi-sphericity of the compression of the fast ignition zones at the pellet's ends. Because the µbunch–µbunch momentum differences are correlated, time-ramped beamline transport elements close after the differential accelerator are used to correct the associated shifts of focal point. Beam neutralization is needed after the differential acceleration until adjacent bunches begin to overlap. Concurrent collapse of each isotope and telescoping of the 16 isotopes cause the current in each beamline to rise

Improvements in streaking nuclear reactors

International Nuclear Information System (INIS)

Pedrick, A.P.

1976-01-01

In this type of reactor atomic nuclei are stripped of their electron shells by heating to form a very high temperature plasma which is passed at high speed through a chamber in which they are forced into contact with a 'wall' formed by a unidirectional stream of photons from continuous laser beams. In this way it should be possible to brush off from the surface of the nuclei protons and neutrons, with release of their binding energy. The energy thus produced can be subjected to much more gentle control than with a fission or fusion reactor. Furthermore, if this concept can be successfully applied to elements of high atomic number which are normally regarded as stable and unfissionable, a vast new source of nuclear energy release will have been made available. It also seems possible that an atomic nucleus might be spun sufficiently in such a reactor to disintegrate it completely into nucleons by simple centrifugal action, with great release of binding energy. The reactor described has a central body with radial ducts through which the nuclei are passed, and a number of lasers are provided whose beams are arranged so that the nuclei are discharged at the cross-over point of two or more laser beams which form a corner at the junction of two or more photon walls. (U.K.)

Radioactive fallout from the Chernobyl nuclear reactor accident

International Nuclear Information System (INIS)

Beiriger, J.M.; Failor, R.A.; Marsh, K.V.; Shaw, G.E.

1987-08-01

This report describes the detection of fallout in the United States from the Chernobyl nuclear reactor accident. As part of its environmental surveillance program, Lawrence Livermore National Laboratory maintained detectors for gamma-emitting radionuclides. Following the reactor accident, additional air filters were set out. Several uncommon isotopes were detected at the time the plume passed into the US

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

International Nuclear Information System (INIS)

Choi, Hangbok

2014-01-01

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

Fuel assembly for gas-cooled nuclear reactors

International Nuclear Information System (INIS)

Yellowlees, J.M.

1976-01-01

A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

Maximization of Transuranic Deep-Burn in High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Kim, Yong Hee; Kim, K. S.; Hong, S. G.; Shim, H. J.; Jo, C. K.; Lee, S. W.

2008-03-01

An optimization study of a single-pass transuranic (TRU) deep burn (DB) has been performed for a block-type modular helium reactor (MHR) proposed. A high-burnup TRU feed vector from light water reactors is considered. For three dimensional equilibrium cores, the performance analysis is done by using the Monte Carlo code McCARD. The core optimization is performed from the viewpoints of the core configuration, fuel management, TRISO fuel specification, and neutron spectrum. With regard to core configuration, two annular cores are investigated in terms of the neutron economy. A conventional radial shuffling scheme of fuel blocks is compared with an axial-only block-shuffling strategy in terms of the fuel bum up and core power distributions. The impact of the kernel size of the TRISO fuel is evaluated, and a diluted kernel, instead of a conventional concentrated kernel, is introduced to maximize the TRU burnup by reducing the self-shielding effects of the TRISO particles. In addition, it is shown that the core power distribution can be effectively controlled by a zoning of the packing fraction of the TRISO fuels. We also have shown that a long-cycle DB-MHR core can be designed by using a two- or three-batch fuel-reloading scheme, at the expense of only a marginal decrease of the TRU discharge bum up. Preliminary safety characteristics of a DBMHR core have been investigated in terms of the temperature coefficients and effective delayed neutron fraction. It has been found that, depending on the fuel management scheme and fuel specifications, the TRU burnup in an optimized DB-MHR core can be over 60% in a single-pass irradiation campaign. In addition, the equilibrium cycle mass balance analyses were also performed for 12 fuel cycles and the impact of TRU deep-bum on the repository was evaluated as well. Additionally, an SFR (Sodium Fast Reactor) fed with DB-MHR spent fuel were designed and characterized

Double-Sided Single-Pass Submerged Arc Welding for 2205 Duplex Stainless Steel

Science.gov (United States)

Luo, Jian; Yuan, Yi; Wang, Xiaoming; Yao, Zongxiang

2013-09-01

The duplex stainless steel (DSS), which combines the characteristics of ferritic steel and austenitic steel, is used widely. The submerged arc welding (SAW) method is usually applied to join thick plates of DSS. However, an effective welding procedure is needed in order to obtain ideal DSS welds with an appropriate proportion of ferrite (δ) and austenite (γ) in the weld zone, particularly in the melted zone and heat-affected zone. This study evaluated the effectiveness of a high efficiency double-sided single-pass (DSSP) SAW joining method for thick DSS plates. The effectiveness of the converse welding procedure, characterizations of weld zone, and mechanical properties of welded joint are analyzed. The results show an increasing appearance and continuous distribution feature of the σ phase in the fusion zone of the leading welded seam. The converse welding procedure promotes the σ phase to precipitate in the fusion zone of leading welded side. The microhardness appears to significantly increase in the center of leading welded side. Ductile fracture mode is observed in the weld zone. A mixture fracture feature appears with a shear lip and tears in the fusion zone near the fusion line. The ductility, plasticity, and microhardness of the joints have a significant relationship with σ phase and heat treatment effect influenced by the converse welding step. An available heat input controlling technology of the DSSP formation method is discussed for SAW of thick DSS plates.

Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

Energy Technology Data Exchange (ETDEWEB)

Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Ho Sik, E-mail: hskim25@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

2015-10-15

Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10{sup −7} in SMRs. • PASS demonstrates fuel coolability (T < 1300 °C) with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10{sup −2} in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10{sup −7}, compared to those of commercial PWRs ∼10{sup −6}–10{sup −1}. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of

Fluidized-bed nuclear reactor

International Nuclear Information System (INIS)

Grimmett, E.S.; Kunze, J.F.

1975-01-01

A reactor vessel containing a fluidized-bed region of particulate material including both a neutron-moderating and a fertile substance is described. A gas flow including fissile material passes through the vessel at a sufficient rate to fluidize the particulate material and at a sufficient density to support a thermal fission reaction within the fluidized-bed region. The high-temperature portion of a heat transfer system is located within the fluidized-bed region of the reactor vessel in direct contact with the fluidized particles. Heat released by fission is thereby transferred at an enhanced rate to a coolant circulating within the heat transfer system. Fission products are continuously removed from the gas flow and supplemental fissile material added during the reactor operation. (U.S.)

Digital control system of advanced reactor

International Nuclear Information System (INIS)

Peng Huaqing; Zhang Rui; Liu Lixin

2001-01-01

This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

Reactor containment

International Nuclear Information System (INIS)

Kawabe, Ryuhei; Yamaki, Rika.

1990-01-01

A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

Coolant clean up system in nuclear reactor

International Nuclear Information System (INIS)

Tajima, Fumio; Iwami, Hiroshi.

1981-01-01

Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

International Nuclear Information System (INIS)

Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

1996-01-01

A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

Mo-99 production on a LEU solution reactor

International Nuclear Information System (INIS)

Brown, R.W.; Thome, L.A.; Khvostionov, V.Y.

2005-01-01

A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO 2 SO 4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)

Safety Evaluation for Packaging for the N Reactor/single pass reactor fuel characterization shipments

International Nuclear Information System (INIS)

Stevens, P.F.

1994-01-01

The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the ChemNuclear CNS 1-13G packaging to ship samples of irradiated fuel elements from the 100 K East and 100 K West basins to the Postirradiation Testing Laboratory (PTL) in support of the spent nuclear fuel characterization effort. It also authorizes the return of the fuel element samples to the 100 K East facility using the same packaging. The CNS 1-13G cask has been-chosen to transport the fuel because it has a Certificate of Compliance (CoC) issued by the US Nuclear Regulatory Commission (NRC) for transporting irradiated oxide and metal fuel in commerce. It is capable of being loaded and offloaded underwater and may be shipped with water in the payload compartment

PLM and the single reactor utility - or how a single reactor utility can face the PLM issues

International Nuclear Information System (INIS)

Ross, M.H.

1994-01-01

Although Gentilly-2 reactor was planned to last for 30 years, its life could be significantly shorter if nothing were done, whereas retubing and refurbishment after, say, 25 years should result in an extension of service life to 45-50 years. In the long run, dimensional changes rather than hydriding may prove to be the pressure tubes' life limiting factor. Hydro Quebec, New Brunswick Power and AECL have an agreement to cooperate in developing a life management program for CANDU-6 reactors. The author expresses the opinion that cost-benefit criteria should be introduced in regulatory decision making. 6 refs., 9 figs

Political-social reactor problems at Berkeley

International Nuclear Information System (INIS)

Little, G.A.

1980-01-01

For better than ten years there was little public notice of the TRIGA reactor at UC-Berkeley. Then: a) A non-student persuaded the Student and Senate to pass a resolution to request Campus Administration to stop operation of the reactor and remove it from campus. b) Presence of the reactor became a campaign-issue in a City Mayoral election. c) Two local residents reported adverse physical reactions before, during, and after a routine tour of the reactor facility. d) The Berkeley City Council began a study of problems associated with radioactive material within the city. e) Friends Of The Earth formally petitioned the NRC to terminate the reactor's license. Campus personnel have expended many man-hours and many pounds of paper in responding to these happenings. Some of the details are of interest, and may be of use to other reactor facilities. (author)

Gold-195m first-pass radionuclide ventriculography, thallium-201 single-photon emission CT, and 12-lead ECG stress testing as a combined procedure

International Nuclear Information System (INIS)

Kipper, S.L.; Ashburn, W.L.; Norris, S.L.; Rimkus, D.S.; Dillon, W.A.

1985-01-01

Graded, sequential, rest/exercise, gold-195m, first-pass ventriculography and thallium-201 (Tl-201) single-photon emission computed tomography (SPECT) were performed simultaneously during a single, electrocardiograph-monitored, bicycle stress test in 24 individuals. The technical aspects and logistics involved in performing this combined radionuclide study are stressed in this preliminary report. Fourteen healthy volunteers each had a normal left ventricular ejection fraction and wall-motion response, along with normal T1-201 perfusion and washout, as determined by both visual and quantitative analysis of the tomographic sections. Each of ten patients with coronary artery disease had at least one abnormality of these parameters. The authors suggest that it is technically feasible to evaluate both cardiac function and myocardial perfusion simultaneously by combing Au-195m ventriculography and Tl-201 SPECT imaging into a single, noninvasive, diagnostic package

Degradation of aqueous phenol solutions by coaxial DBD reactor

Science.gov (United States)

Dojcinovic, B. P.; Manojlovic, D.; Roglic, G. M.; Obradovic, B. M.; Kuraica, M. M.; Puric, J.

2008-07-01

Solutions of 2-chlorophenol, 4-chlorophenol and 2,6-dichlorophenol in bidistilled and water from the river Danube were treated in plasma reactor. In this reactor, based on coaxial dielectric barrier discharge at atmospheric pressure, plasma is formed over a thin layer of treated water. After one pass through the reactor, starting chlorophenols concentration of 20 mg/l was diminished up to 95 %. Kinetics of the chlorophenols degradation was monitored by High Pressure Liquid Chromatography method (HPLC).

[Mobile single-pass batch hemodialysis system in intensive care medicine. Reduction of costs and workload in renal replacement therapy].

Science.gov (United States)

Hopf, H-B; Hochscherf, M; Jehmlich, M; Leischik, M; Ritter, J

2007-07-01

This paper describes the introduction of a single-pass batch hemodialysis system for renal replacement therapy in a 14 bed intensive care unit. The goals were to reduce the workload of intensive care unit physicians using an alternative and simpler method compared to continuous veno-venous hemodiafiltration (CVVHDF) and to reduce the costs of hemofiltrate solutions (80,650 EUR per year in our clinic in 2005). We describe and evaluate the process of implementation of the system as well as the achieved and prospective savings. We conclude that a close cooperation of all participants (physicians, nurses, economists, technicians) of a hospital can achieve substantial benefits for patients and employees as well as reduce the economic burden of a hospital.

Coolant cleanup method in a nuclear reactor

International Nuclear Information System (INIS)

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

1983-01-01

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

A model to describe the performance of the UASB reactor.

Science.gov (United States)

Rodríguez-Gómez, Raúl; Renman, Gunno; Moreno, Luis; Liu, Longcheng

2014-04-01

A dynamic model to describe the performance of the Upflow Anaerobic Sludge Blanket (UASB) reactor was developed. It includes dispersion, advection, and reaction terms, as well as the resistances through which the substrate passes before its biotransformation. The UASB reactor is viewed as several continuous stirred tank reactors connected in series. The good agreement between experimental and simulated results shows that the model is able to predict the performance of the UASB reactor (i.e. substrate concentration, biomass concentration, granule size, and height of the sludge bed).

Reactor container spray device

International Nuclear Information System (INIS)

Yanai, Ryoichi.

1980-01-01

Purpose: To enable decrease in the heat and the concentration of radioactive iodine released from the reactor vessel into the reactor container in the spray device of BWR type reactors. Constitution: A plurality of water receiving trays are disposed below the spray nozzle in the dry well and communicated to a pressure suppression chamber by way of drain pipeways passing through a diaphragm floor. When the recycling system is ruptured and coolants in the reactor vessel and radioactive iodine in the reactor core are released into the dry well, spray water is discharged from the spray nozzle to eliminate the heat and the radioactive iodine in the dry well. In this case, the receiving trays collect the portions of spray water whose absorption power for the heat and radioactive iodine is nearly saturated and falls them into the pool water of the pressure suppression chamber. Consequently, other portions of the spray water that still possess absorption power can be jetted with no hindrance, to increase the efficiency for the removal of the heat and iodine of the spray droplets. (Horiuchi, T.)

Single-step syngas-to-distillates (S2D) process based on biomass-derived syngas--a techno-economic analysis.

Science.gov (United States)

Zhu, Yunhua; Jones, Susanne B; Biddy, Mary J; Dagle, Robert A; Palo, Daniel R

2012-08-01

This study compared biomass gasification based syngas-to-distillate (S2D) systems using techno-economic analysis (TEA). Three cases, state of technology (SOT), goal, and conventional, were compared in terms of performance and cost. The SOT case represented the best available experimental results for a process starting with syngas using a single-step dual-catalyst reactor for distillate generation. The conventional case mirrored a conventional two-step S2D process consisting of separate syngas-to-methanol and methanol-to-gasoline (MTG) processes. The goal case assumed the same performance as the conventional, but with a single-step S2D technology. TEA results revealed that the SOT was more expensive than the conventional and goal cases. The SOT case suffers from low one-pass yield and high selectivity to light hydrocarbons, both of which drive up production cost. Sensitivity analysis indicated that light hydrocarbon yield and single pass conversion efficiency were the key factors driving the high cost for the SOT case. Copyright © 2012 Elsevier Ltd. All rights reserved.

FLUID MODERATED REACTOR

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1957-10-22

A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

Passing and Catching in Rugby.

Science.gov (United States)

Namudu, Mike M.

This booklet contains the fundamentals for rugby at the primary school level. It deals primarily with passing and catching the ball. It contains instructions on (1) holding the ball for passing, (2) passing the ball to the left--standing, (3) passing the ball to the left--running, (4) making a switch pass, (5) the scrum half's normal pass, (6) the…

A double-pass interferometer for measurement of dimensional changes

International Nuclear Information System (INIS)

Ren, Dongmei; Lawton, K M; Miller, J A

2008-01-01

A double-pass interferometer was developed for measuring dimensional changes of materials in a nanoscale absolute interferometric dilatometer. This interferometer realized the double-ended measurement of a sample using a single-detection double-pass interference system. The nearly balanced design, in which the measurement beam and the reference beam have equal optical path lengths except for the path difference caused by the sample itself, makes this interferometer have high stability, which is verified by the measurement of a quasi-zero-length sample. The preliminary experiments and uncertainty analysis show that this interferometer should be able to measure dimensional changes with characteristic uncertainty at the nanometer level

Reactor outage schedule (tentative)

Energy Technology Data Exchange (ETDEWEB)

Walton, R.P.

1969-11-01

This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

Reactor outage schedule (tentative)

Energy Technology Data Exchange (ETDEWEB)

Walton, R.P.

1969-10-01

This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

Reactor outage schedule (tentative)

Energy Technology Data Exchange (ETDEWEB)

Walton, R.P.

1969-10-15

This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

Reactor outage schedule (tentative)

Energy Technology Data Exchange (ETDEWEB)

Walton, R.P.

1969-09-15

This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

Reactor outage schedule (tentative)

Energy Technology Data Exchange (ETDEWEB)

Walton, R.P.

1969-12-15

This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

Reactor outage schedule (tentative)

Energy Technology Data Exchange (ETDEWEB)

Walton, R.P.

1969-12-01

This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

Water injection device for reactor container

International Nuclear Information System (INIS)

Sakaki, Isao.

1996-01-01

A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

Sea ice local surface topography from single-pass satellite InSAR measurements: a feasibility study

Directory of Open Access Journals (Sweden)

W. Dierking

2017-08-01

Full Text Available Quantitative parameters characterizing the sea ice surface topography are needed in geophysical investigations such as studies on atmosphere–ice interactions or sea ice mechanics. Recently, the use of space-borne single-pass interferometric synthetic aperture radar (InSAR for retrieving the ice surface topography has attracted notice among geophysicists. In this paper the potential of InSAR measurements is examined for several satellite configurations and radar frequencies, considering statistics of heights and widths of ice ridges as well as possible magnitudes of ice drift. It is shown that, theoretically, surface height variations can be retrieved with relative errors  ≤  0.5 m. In practice, however, the sea ice drift and open water leads may contribute significantly to the measured interferometric phase. Another essential factor is the dependence of the achievable interferometric baseline on the satellite orbit configurations. Possibilities to assess the influence of different factors on the measurement accuracy are demonstrated: signal-to-noise ratio, presence of a snow layer, and the penetration depth into the ice. Practical examples of sea surface height retrievals from bistatic SAR images collected during the TanDEM-X Science Phase are presented.

Multi-pass TIG welding process: simulating thermal SS304

International Nuclear Information System (INIS)

Harinadh, Vemanaboina; Akella, S.; Buddu, Ramesh Kumar; Edision, G.

2015-01-01

Welding is basic requirement in the construction of nuclear reactors, power plants and structural components development. A basic studies on various aspects of the welding is essential to ensure the stability and structural requirement conditions. The present study explored the thermo-mechanical analysis of the multipass welds of austenitic stainless steels which are widely used in fusion and fission reactor components development. A three-dimensional (3D) finite element model is developed to investigate thermally induced stress field during TIG welding process for SS304 material. The transient thermal analysis is performed to obtain the temperature history, which then is applied to the mechanical (stress) analysis. The present thermal analysis is conducted using element type DC3D8. This element type has a three dimensional thermal conduction capability and eight nodes. The 6 mm thick plated is welded with six numbers of passes. The geometry and meshed model with tetrahedral shape with volume sweep. The analysis is on TIG welding process using 3D-weld interface plug-in on ABAQUS-6.14. The results are reported in the present paper

The fast breeder reactor

International Nuclear Information System (INIS)

Keck, O.

1984-01-01

Nowadays the fast-breeder reactor is a negative symbol of advanced technology which is getting out of control and, due to its complexity, is incomprehensible for politicians and therefore by-passes the established order. The author lists the most important decisions over state aid to the fast-breeder-reactors up until the mid-seventies and uses documents from the appropriate advisory bodies as reference. He was also aided by interviews with those directly involved with the project. The empirical facts forces us to discard our traditional view of the relationship between state and industry with regard to advanced technology. The author explains that it is impossible to find any economic value in the fast-breeder reactor. The insight gained through this project allows him to draw conclusions which apply to all aspects of state aid to advanced technology. (orig.) [de

Stability of an anaerobic single reactor filled with dolomitic limestone with increased organic load of sugarcane

Directory of Open Access Journals (Sweden)

Maria Magdalena Ribas Döll

2017-12-01

Full Text Available The anaerobic single-stage reactor was evaluated to treat vinasse and to evaluate its stability. This bench reactor was filled with dolomitic limestone with a horizontal plug flow to simulate a drainage channel. The experiment lasted 129 days while the reactor was submitted to different applied organic concentrations (chronologically applied: 3.0; 5.0; 12.0; 9.0 and 7.5 g L-1 as COD, chemical oxygen demand. COD removals were 50% and 9% with 3.0 and 7.5 g L-1, respectively. With 12.0 g L-1, reactor efficiency increased to 33%, with an abrupt drop to 3% on the 84th day. Therefore, in order to avoid reactor collapse, a remedial measure was necessary. The system remained in batch without feeding for 19 days (from the 85th to the 104th day with 9.0 g L-1. Afterwards, it was observed that the performance of the system tended to stabilize, reaching 47% with 7.5 g L-1 in the 118th day. At the end of the experiment, the potassium content of the wastewater decreased from 800 mg L-1 to 594 mg L-1 (on an average 25% and calcium and magnesium increased within the reactor liquor. The dissolution of the limestone inside the liquor reactor probably caused this result. After the treatment with limestone, the average pH value of the effluent increased from 4.9 to over 6.0 in all organic concentrations. It could be concluded that the reactor filled with dolomitic limestone in these operational conditions assured a low efficiency in COD removal, potassium reduction, increasing values of pH, alkalinity, calcium and magnesium. The instability was observed when there was increase in organic load to 12 g L-1 with subsequent recovery.

Device for controlling water supply to nuclear reactor

International Nuclear Information System (INIS)

Iwasaki, Toshio.

1974-01-01

Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the signals from a supplied water rate limiting signal generator generating signals for indicating whether one or two water supply pumps are operated. A low value preferential circuit passes the lower of the values generated from the selection circuit and the adder. The selection circuit receives a recirculation pump condition signal and selects either one of the signals from the supplied water flow rate limiting signal generator operated at high speed or low speed. A high value preferential circuit passes the higher value

Loop-type FBR reactor

International Nuclear Information System (INIS)

Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

1996-01-01

The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

Some local dilution transient in a pressurized water reactor

International Nuclear Information System (INIS)

Jacobson, S.

1989-01-01

Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

Application of the single-channel continuous synthesis method to criticity and power distribution calculations in thermal reactors

International Nuclear Information System (INIS)

Medrano Asensio, Gregorio.

1976-06-01

A detailed power distribution calculation in a large power reactor requires the solution of the multigroup 3D diffusion equations. Using the finite difference method, this computation is too expensive to be performed for design purposes. This work is devoted to the single channel continous synthesis method: the choice of the trial functions and the determination of the mixing functions are discussed in details; 2D and 3D results are presented. The method is applied to the calculation of the IAEA ''Benchmark'' reactor and the results obtained are compared with a finite element resolution and with published results [fr

Field trial of a fast single-pass transmit-receive probe during Gentilly II steam generator tube inspection

International Nuclear Information System (INIS)

Obrutsky, L.; Cantin, M.; Renaud, J.; Cecco, V.; Lakhan, R.; Sullivan, S.

2000-01-01

A new generation of transmit-receive single-pass probes, denoted as C6 or X probe, was field tested during the Gentilly II, 2000 steam generator tube inspection. This probe has a performance equivalent to rotating probes and can be used for tubesheet and full-length inspection at an inspection speed equivalent to that of bobbin probes. Existing C3 transmit-receive probes have been demonstrated to be effective in detecting circumferential cracks. The C5 probe can detect both circumferential and axial cracks and volumetric defects but cannot discriminate between them. The C6 probe expands on the capabilities of both probes in a single probe head. It can simultaneously detect and discriminate between circumferential and axial cracks to satisfy different plugging criteria. It has excellent coverage, good defect detectability, and improved sizing and characterization. Probe data is displayed in C-scan format so that the amount of data to be analyzed is similar to rotating probes. The C6 probe will significantly decrease inspection time and the need for re-inspection and tube pulling. This paper describes the advantages of the probe and demonstrates its capabilities employing signals from tube samples with calibration flaws and laboratory induced cracks. It shows the results from the field trial of the probe at Gentilly II and describes the instrumentation, hardware and software used for the inspection. (author)

Field trial of a fast single-pass transmit-receive probe during Gentilly II steam generator tube inspection

International Nuclear Information System (INIS)

Obrutsky, L.; Cantin, M.; Renaud, J.; Cecco, V.; Lakhan, R.; Sullivan, S.

2000-01-01

A new generation of transmit-receive single-pass probes, denoted as C6 or X probe, was field-tested during the Gentilly II, 2000 steam generator tube inspection. This probe has a performance equivalent to rotating probes and can be used for tubesheet and full-length inspection at an inspection speed equivalent to that of bobbin probes. Existing C3 transmit-receive probes have been demonstrated to be effective in detecting circumferential cracks. The C5 probe can detect both circumferential and axial cracks and volumetric defects but cannot discriminate between them. The C6 probe expands on the capabilities of both probes in a single probe head. It can simultaneously detect and discriminate between circumferential and axial cracks to satisfy different plugging criteria. It has excellent coverage, good defect detectability, and improved sizing and characterization. Probe data is displayed in C-scan format so that the amount of data to be analyzed is similar to rotating probes. The C6 probe will significantly decrease inspection time and the need for re-inspection and tube pulling. This paper describes the advantages of the probe and demonstrates its capabilities employing signals from tube samples with calibration flaws and laboratory induced cracks. It shows the results from the field trial of the probe at Gentilly II and describes the instrumentation, hardware and software used for the inspection. (author)

Buffels (Wes), Elsies, Sir Lowry's pass, Steenbras and Buffels (Oos)

CSIR Research Space (South Africa)

Heinecken, TJE

1982-06-01

Full Text Available This report is a synthesis of all available information on five of the smaller rivers discharging in to False bay combined as a single volume. The rivers dealt with are the Buffels (Wes), Elsies, Sir Lowry' Pass, Steenbras and Buffels (Oos). False...

Green synthesis of isopropyl myristate in novel single phase medium Part II: Packed bed reactor (PBR) studies.

Science.gov (United States)

Vadgama, Rajeshkumar N; Odaneth, Annamma A; Lali, Arvind M

2015-12-01

Isopropyl myristate is a useful functional molecule responding to the requirements of numerous fields of application in cosmetic, pharmaceutical and food industry. In the present work, lipase-catalyzed production of isopropyl myristate by esterification of myristic acid with isopropyl alcohol (molar ratio of 1:15) in the homogenous reaction medium was performed on a bench-scale packed bed reactors, in order to obtain suitable reaction performance data for upscaling. An immobilized lipase B from Candida antartica was used as the biocatalyst based on our previous study. The process intensification resulted in a clean and green synthesis process comprising a series of packed bed reactors of immobilized enzyme and water dehydrant. In addition, use of the single phase reaction system facilitates efficient recovery of the product with no effluent generated and recyclability of unreacted substrates. The single phase reaction system coupled with a continuous operating bioreactor ensures a stable operational life for the enzyme.

Green synthesis of isopropyl myristate in novel single phase medium Part II: Packed bed reactor (PBR studies

Directory of Open Access Journals (Sweden)

Rajeshkumar N. Vadgama

2015-12-01

Full Text Available Isopropyl myristate is a useful functional molecule responding to the requirements of numerous fields of application in cosmetic, pharmaceutical and food industry. In the present work, lipase-catalyzed production of isopropyl myristate by esterification of myristic acid with isopropyl alcohol (molar ratio of 1:15 in the homogenous reaction medium was performed on a bench-scale packed bed reactors, in order to obtain suitable reaction performance data for upscaling. An immobilized lipase B from Candida antartica was used as the biocatalyst based on our previous study. The process intensification resulted in a clean and green synthesis process comprising a series of packed bed reactors of immobilized enzyme and water dehydrant. In addition, use of the single phase reaction system facilitates efficient recovery of the product with no effluent generated and recyclability of unreacted substrates. The single phase reaction system coupled with a continuous operating bioreactor ensures a stable operational life for the enzyme.

Single-reactor process for producing liquid-phase organic compounds from biomass

Science.gov (United States)

Dumesic, James A [Verona, WI; Simonetti, Dante A [Middleton, WI; Kunkes, Edward L [Madison, WI

2011-12-13

Disclosed is a method for preparing liquid fuel and chemical intermediates from biomass-derived oxygenated hydrocarbons. The method includes the steps of reacting in a single reactor an aqueous solution of a biomass-derived, water-soluble oxygenated hydrocarbon reactant, in the presence of a catalyst comprising a metal selected from the group consisting of Cr, Mn, Fe, Co, Ni, Cu, Mo, Tc, Ru, Rh, Pd, Ag, W, Re, Os, Ir, Pt, and Au, at a temperature, and a pressure, and for a time sufficient to yield a self-separating, three-phase product stream comprising a vapor phase, an organic phase containing linear and/or cyclic mono-oxygenated hydrocarbons, and an aqueous phase.

MEANS FOR SHIELDING AND COOLING REACTORS

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

Multigroup calculation of criticality and power distribution in a two-pass fast spectrum cermet-fueled reactor

International Nuclear Information System (INIS)

Anghaie, S.; Feller, G.J.; Peery, S.D.; Parsley, R.C.

1992-01-01

The advanced propulsion group at Pratt ampersand Whitney has developed a nuclear thermal rocket concept, the XNR2000, for use on lunear, Mars, and deep-space planetary missions. The XNR2000 engine is powered by a fast spectrum cermet-fueled nuclear reactor that heats up hydrogen propellant to a maximum of 2850 K. An expander cycle is used to deliver 12 kg/s hydrogen to the core, producing 25,000 lb f thrust at 944 s of specific impulse. The reactor comprises a beryllium-reflected outer annulus core and an inner core with the hydrogen propellant entering from the bottom of the outer core and exiting from the bottom part of the inner core to the thrust chamber. Both the outer and inner cores are loaded with prismatic cermet fuel elements. The baseline XNR2000 reactor core consists of 90 fuel elements in the outer core and 61 in the inner core, arranged in the pattern. This paper focuses on the neutronic analysis of the baseline XNR2000 reactor

Reactor water clean-up device

International Nuclear Information System (INIS)

Sawa, Toshio; Takahashi, Sankichi; Takashima, Yoshie.

1983-01-01

Purpose: To efficiently eliminate radioactive materials such as iron oxide and cobalt ions with less heat loss by the use of an electrode assembly applied with a direct current. Constitution: In a reactor water clean-up device adapted to pass reactor water through an electrode assembly comprising at least a pair of anode and cathode applied with a direct current to eliminate various types of ions contained in the reactor water by way of the electrolysis or charge neutralization at the anode, the cathode is constituted with a corrosion resistant grid-like or porous metal plate and a layer to the upper portion of the metal plate filled with a plurality of metal spheres of about 1 - 5 mm diameter, and the anode is made of insoluble porous or spirally wound metal material. (Seki, T.)

Neutron noise in nuclear reactors

International Nuclear Information System (INIS)

Blaquiere, A.; Pachowska, R.

1961-06-01

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [fr

Separated type nuclear superheating reactor

International Nuclear Information System (INIS)

Hida, Kazuki.

1993-01-01

In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

Device for controlling a recirculation flow in a reactor

International Nuclear Information System (INIS)

Shida, Toichi; Tohei, Kazushige; Hirose, Masao; Nakamura, Hideo.

1976-01-01

Object: To provide an emergency cut-off valve in a recirculation system in a reactor to control the recirculation at the time of turbine trip or load cut-off, thereby relieving excessive increase in heat output of fuel. Structure: A recirculation pump is driven through a recirculation pump motor by an AC generator, which is driven by a driving motor through a fluid coupling, so that reactor water passes the emergency cut-off valve and recirculation flow stop valve and then passes a jet pump into the core. At the time of turbine trip or load cut-off, the emergency cut-off valve is closed by a hydraulic circuit, whereby core flow is merely decreased by 20 to 30% in a short period of time to restrain excessive increase in heat output. (Yoshino, Y.)

Comparison of the gravimetric, phenol red, and 14C-PEG-3350 methods to determine water absorption in the rat single-pass intestinal perfusion model

OpenAIRE

Sutton, Steven C.; Rinaldi, M. T. S.; Vukovinsky, K. E.

2001-01-01

This study was undertaken to determine whether the gravimetric method provided an accurate measure of water flux correction and to compare the gravimetric method with methods that employ nonabsorbed markers (eg, phenol red and 14C-PEG-3350). Phenol red, 14C-PEG-3350, and 4-[2-[[2-(6-amino-3-pyridinyl)-2-hydroxyethyl]amino]ethoxy]-methyl ester, (R)-benzene acetic acid (Compound I) were co-perfused in situ through the jejunum of 9 anesthetized rats (single-pass intestinal perfusion [SPIP]). Wat...

The integral fast reactor

International Nuclear Information System (INIS)

Till, C.E.

1987-01-01

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

An operational protocol for facilitating start-up of single-stage autotrophic nitrogen-removing reactors based on process stoichiometry

DEFF Research Database (Denmark)

Mutlu, Ayten Gizem; Vangsgaard, Anna Katrine; Sin, Gürkan

2013-01-01

Start-up and operation of single-stage nitritation–anammox sequencing batch reactors (SBRs) for completely autotrophic nitrogen removal can be challenging and far from trivial. In this study, a step-wise procedure is developed based on stoichiometric analysis of the process performance from...

Comparison of cryogenic low-pass filters

Science.gov (United States)

Thalmann, M.; Pernau, H.-F.; Strunk, C.; Scheer, E.; Pietsch, T.

2017-11-01

Low-temperature electronic transport measurements with high energy resolution require both effective low-pass filtering of high-frequency input noise and an optimized thermalization of the electronic system of the experiment. In recent years, elaborate filter designs have been developed for cryogenic low-level measurements, driven by the growing interest in fundamental quantum-physical phenomena at energy scales corresponding to temperatures in the few millikelvin regime. However, a single filter concept is often insufficient to thermalize the electronic system to the cryogenic bath and eliminate spurious high frequency noise. Moreover, the available concepts often provide inadequate filtering to operate at temperatures below 10 mK, which are routinely available now in dilution cryogenic systems. Herein we provide a comprehensive analysis of commonly used filter types, introduce a novel compact filter type based on ferrite compounds optimized for the frequency range above 20 GHz, and develop an improved filtering scheme providing adaptable broad-band low-pass characteristic for cryogenic low-level and quantum measurement applications at temperatures down to few millikelvin.

Comparison of cryogenic low-pass filters.

Science.gov (United States)

Thalmann, M; Pernau, H-F; Strunk, C; Scheer, E; Pietsch, T

2017-11-01

Low-temperature electronic transport measurements with high energy resolution require both effective low-pass filtering of high-frequency input noise and an optimized thermalization of the electronic system of the experiment. In recent years, elaborate filter designs have been developed for cryogenic low-level measurements, driven by the growing interest in fundamental quantum-physical phenomena at energy scales corresponding to temperatures in the few millikelvin regime. However, a single filter concept is often insufficient to thermalize the electronic system to the cryogenic bath and eliminate spurious high frequency noise. Moreover, the available concepts often provide inadequate filtering to operate at temperatures below 10 mK, which are routinely available now in dilution cryogenic systems. Herein we provide a comprehensive analysis of commonly used filter types, introduce a novel compact filter type based on ferrite compounds optimized for the frequency range above 20 GHz, and develop an improved filtering scheme providing adaptable broad-band low-pass characteristic for cryogenic low-level and quantum measurement applications at temperatures down to few millikelvin.

Determination of reactor parameters by single rod experiments

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Zdravkovic, Z; Ivkovic, M [Department of Reactor Physics and Dynamics, Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

1969-07-01

A method is developed for the experimental determination of reactor parameters by using an isolated fuel element. The method is based on the consideration of the fuel element as the source and sink of neutrons when placed in a constant neutron field. By measuring the perturbation of the original field produced by insertion of the test fuel element it was possible to determine the fuel element parameters defined by the heterogeneous reactor theory of Feinberg and Galanin as thermal neutron absorption constant {gamma}, and neutron multiplication constant {eta}. Statistical error for one series of measurement amount to 2% in the values of {eta} and {gamma}. The developed method was intended for the analysis of the nuclear characteristics of the fuel element in the stage of its construction and development for a given reactor system. (author)

Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

International Nuclear Information System (INIS)

1995-07-01

The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D ampersand D) and to reduce the cost of maintaining the facilities prior to D ampersand D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor's fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered

Assessment of PASS Effectiveness under Severe Accidents in Nuclear Power Plants

International Nuclear Information System (INIS)

Choi, Yu Jung; Lee, Sung Bok; Kim, Hyeong Taek; Lee, Jin Yong

2008-01-01

Following the accident at Three Mile Island Unit 2 (TMI-2) on March 28, 1979, the USNRC formed a lessons-learned Task Force to identify and evaluate safety concerns originating with the TMI-2 accident. NUREG-0578 documented the results of the task force effort. One of the recommendations of the task force was for licensees to upgrade the capability to obtain samples from the reactor coolant system and containment atmosphere under high radioactivity conditions and to provide the capability for chemical and spectral analyses of high-level samples on site. NUREG-0737 contained the details of the TMI recommendations that were to be implemented by the licensees. Additional criteria for post accident sampling system(PASS) were issued by Regulatory Guide 1.97. As the results, PASS has been installed on nuclear power plants(NPPs) in Korea as well as United States. However, significant improvements have been achieved since the TMI-2 accident in the areas of understanding risks associated with nuclear plant operations and developing better strategies for managing the response to potential severe accidents at NPPs. Thus, the requirements for PASS have been re-evaluated in some reports. According to the reports, the samples and measurements from PASS do not contribute significantly to emergency management response to severe accidents due to the long analyzing time, 3 hours. Hence, this paper focused on the development of the quantitative analysis methodology to analyze the sequence of the severe accident in Yonggwang nuclear power plants (YGN) and presented the results of the analysis according to the developed methodology

Pulse propagation in a two-pass optical amplifier with arbitrary laser beams overlap

Directory of Open Access Journals (Sweden)

AH Farahbod

2011-09-01

Full Text Available An analytical model for two-pass optical amplifier with arbitrary beams overlap has been developed which generalized the classical theory of Frantz-Nodvik for single pass amplifier. The effect of counterpropagating beams on gain and output energy fluence included in the model. Moreover, the appropriate limiting relations for two special cases of weak input signal and saturation state of the amplifier gain have been derived. The results indicate that for complete beams overlap, the gain and output energy have the least values. The model predictions are consistent with experimental observations and exact analytical model for two-pass amplifier when beam propagation paths are coincided.

An operation protocol for facilitating start-up of single-stage autotrophic nitrogen removing reactors based on process stoichiometry

DEFF Research Database (Denmark)

Mutlu, A. Gizem; Vangsgaard, Anna Katrine; Sin, Gürkan

2012-01-01

Start-up and operation of single-stage nitritation/anammox reactor employing complete autotrophic nitrogen can be difficult. Keeping the performance criteria and monitoring the microbial community composition may not be easy or fast enough to take action on time. In this study, a control strategy...

Improvement to reactor vessel

International Nuclear Information System (INIS)

1974-01-01

The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

Determining the Optimal Number of Core Needle Biopsy Passes for Molecular Diagnostics.

Science.gov (United States)

Hoang, Nam S; Ge, Benjamin H; Pan, Lorraine Y; Ozawa, Michael G; Kong, Christina S; Louie, John D; Shah, Rajesh P

2018-03-01

The number of core biopsy passes required for adequate next-generation sequencing is impacted by needle cut, needle gauge, and the type of tissue involved. This study evaluates diagnostic adequacy of core needle lung biopsies based on number of passes and provides guidelines for other tissues based on simulated biopsies in ex vivo porcine organ tissues. The rate of diagnostic adequacy for pathology and molecular testing from lung biopsy procedures was measured for eight operators pre-implementation (September 2012-October 2013) and post-implementation (December 2013-April 2014) of a standard protocol using 20-gauge side-cut needles for ten core biopsy passes at a single academic hospital. Biopsy pass volume was then estimated in ex vivo porcine muscle, liver, and kidney using side-cut devices at 16, 18, and 20 gauge and end-cut devices at 16 and 18 gauge to estimate minimum number of passes required for adequate molecular testing. Molecular diagnostic adequacy increased from 69% (pre-implementation period) to 92% (post-implementation period) (p < 0.001) for lung biopsies. In porcine models, both 16-gauge end-cut and side-cut devices require one pass to reach the validated volume threshold to ensure 99% adequacy for molecular characterization, while 18- and 20-gauge devices require 2-5 passes depending on needle cut and tissue type. Use of 20-gauge side-cut core biopsy needles requires a significant number of passes to ensure diagnostic adequacy for molecular testing across all tissue types. To ensure diagnostic adequacy for molecular testing, 16- and 18-gauge needles require markedly fewer passes.

Recent palladium membrane reactor development at the tritium systems test assembly

International Nuclear Information System (INIS)

Scott, W.R.; Birdsell, S.A.; Wilhelm, R.C.

1995-01-01

The palladium membrane reactor (PMR) is being investigated as a means for recovering hydrogen isotopes (including tritium) from compounds such as water and methane. Previous work with protiated water and methane showed that this device can be used to obtain high hydrogen recovery efficiencies using a single processing pass and with essentially no waste production. With these successful proof-of-principle results completed, recent work has focused on PMR development. This included studies of various geometries and testing with tritium. The results, which are reported here, have led to a better understanding of the PMR and will lead to the ultimate goal of building a production PMR and putting it into practical tritium processing service. 3 refs., 5 figs., 1 tab

Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

2002-08-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

Science.gov (United States)

Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

2015-12-01

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

International Nuclear Information System (INIS)

Hamann, S.; Röpcke, J.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.

2015-01-01

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH 4 , C 2 H 2 , HCN, and NH 3 ). With the help of OES, the rotational temperature of the screen plasma could be determined

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

Energy Technology Data Exchange (ETDEWEB)

Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

2015-12-15

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

Requirements of coolants in nuclear reactors

International Nuclear Information System (INIS)

Abass, O. A. M.

2014-11-01

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

Self operation type reactor control device

International Nuclear Information System (INIS)

Saito, Makoto; Gunji, Minoru.

1990-01-01

A boiling-requefication chamber containing transporting materials having somewhat higher boiling point that the usual reactor operation temperature and liquid neutron absorbers having a boiling point sufficiently higher than that of the transporting materials is disposed near the coolant exit of a fuel assembly and connected with a tubular chamber in the reactor core with a moving pipe at the bottom. Since the transporting materials in the boiling-requefication chamber is boiled and expanded by heating, the liquid neutron absorbers are introduced passing through the moving pipe into the cylindrical chamber to control the nuclear reactions. When the temperature is lowered by the control, the transporting materials are liquefied to contract the volume and the liquid neutron absorbers in the cylindrical chamber are returned passing through the moving tube into the boiling-liquefication chamber to make the nuclear reaction vigorous. Thus, self-operation type power conditioning and power stopping are enabled not by way of control rods and not requiring external control, to prevent scram failure or misoperation. (N.H.)

CMOS-based carbon nanotube pass-transistor logic integrated circuits

Science.gov (United States)

Ding, Li; Zhang, Zhiyong; Liang, Shibo; Pei, Tian; Wang, Sheng; Li, Yan; Zhou, Weiwei; Liu, Jie; Peng, Lian-Mao

2012-01-01

Field-effect transistors based on carbon nanotubes have been shown to be faster and less energy consuming than their silicon counterparts. However, ensuring these advantages are maintained for integrated circuits is a challenge. Here we demonstrate that a significant reduction in the use of field-effect transistors can be achieved by constructing carbon nanotube-based integrated circuits based on a pass-transistor logic configuration, rather than a complementary metal-oxide semiconductor configuration. Logic gates are constructed on individual carbon nanotubes via a doping-free approach and with a single power supply at voltages as low as 0.4 V. The pass-transistor logic configurarion provides a significant simplification of the carbon nanotube-based circuit design, a higher potential circuit speed and a significant reduction in power consumption. In particular, a full adder, which requires a total of 28 field-effect transistors to construct in the usual complementary metal-oxide semiconductor circuit, uses only three pairs of n- and p-field-effect transistors in the pass-transistor logic configuration. PMID:22334080

Coupled neutronic/thermal-hydraulic analysis of the HPLWR three pass core

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2008-01-01

The High Performance Light Water Reactor is an innovative Gen-IV reactor cooled and moderated with water at supercritical pressure. The three pass core concept has been proposed to reduce peaking factors, i.e. hot-channel effects, and it further increases the core heterogeneity, which is mainly due to pronounced water density reduction. For this kind of nuclear reactor, the significant feedbacks - which exist between the properties of the components and the power generation rate - can not be neglected and require a coupled Neutronic/Thermal-Hydraulic analysis even for steady state conditions. The main goal of this paper is to present the developed tool for coupled analyses of the HPLWR. Two state-of-the-art codes have been chosen for Thermal-Hydraulic and Neutronic core analyses, namely TRACE and ERANOS, and they have been coupled with in an iterative procedure in which they are run in series until a steady state condition has been reached. In the simplifying assumptions of uniform enrichment distribution, zero burn-up and ignoring the effect of the control rods, the obtained steady state condition will be discussed and a core power map, flow rate redistribution as well as water and fuel temperature variations will be presented. (author)

Nonlinear Dynamic Model of Power Plants with Single-Phase Coolant Reactors

International Nuclear Information System (INIS)

Vollmer, H.

1968-12-01

The traditional way of developing dynamic models for a specific nuclear power plant and for specific purpose seems rather uneconomical, as much of the information often can not be utilized if the plant design or the required accuracy of the calculation is desired to be changed. It is therefore suggested that the model development may be made more systematic, general and flexible by - applying the 'box of bricks' system, where the main components of a nuclear power plant are treated separately and combined afterwards according to a given flow scheme, - a dynamic determination of the components which is as general as possible without taking into account those details which have a minor influence on the overall dynamics, - providing approximations of the more rigorous solution sufficient to meet the user s requirements on accuracy, - proper use of computers. A dynamic model for single-phase coolant reactor plants is established along these lines. By separation of the nonlinear and linear parts of the system, application of Laplace transformation and proper approximations, and the use of a hybrid computer it seems possible to determine the (nonlinear) dynamic behaviour of such a plant for perturbations which are not so large that phase changes of physical parameters occur, e. g. fuel does not melt. The model is applied to a steam cooled fast reactor power plant

Nonlinear Dynamic Model of Power Plants with Single-Phase Coolant Reactors

Energy Technology Data Exchange (ETDEWEB)

Vollmer, H

1968-12-15

The traditional way of developing dynamic models for a specific nuclear power plant and for specific purpose seems rather uneconomical, as much of the information often can not be utilized if the plant design or the required accuracy of the calculation is desired to be changed. It is therefore suggested that the model development may be made more systematic, general and flexible by - applying the 'box of bricks' system, where the main components of a nuclear power plant are treated separately and combined afterwards according to a given flow scheme, - a dynamic determination of the components which is as general as possible without taking into account those details which have a minor influence on the overall dynamics, - providing approximations of the more rigorous solution sufficient to meet the user s requirements on accuracy, - proper use of computers. A dynamic model for single-phase coolant reactor plants is established along these lines. By separation of the nonlinear and linear parts of the system, application of Laplace transformation and proper approximations, and the use of a hybrid computer it seems possible to determine the (nonlinear) dynamic behaviour of such a plant for perturbations which are not so large that phase changes of physical parameters occur, e. g. fuel does not melt. The model is applied to a steam cooled fast reactor power plant.

Nuclear reactor containment device

International Nuclear Information System (INIS)

Ichiki, Tadaharu.

1980-01-01

Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

Uranium Release from Acidic Weathered Hanford Sediments: Single-Pass Flow-Through and Column Experiments.

Science.gov (United States)

Wang, Guohui; Um, Wooyong; Wang, Zheming; Reinoso-Maset, Estela; Washton, Nancy M; Mueller, Karl T; Perdrial, Nicolas; O'Day, Peggy A; Chorover, Jon

2017-10-03

The reaction of acidic radioactive waste with sediments can induce mineral transformation reactions that, in turn, control contaminant fate. Here, sediment weathering by synthetic uranium-containing acid solutions was investigated using bench-scale experiments to simulate waste disposal conditions at Hanford's cribs (Hanford, WA). During acid weathering, the presence of phosphate exerted a strong influence over uranium mineralogy and a rapidly precipitated, crystalline uranium phosphate phase (meta-ankoleite [K(UO 2 )(PO 4 )·3H 2 O]) was identified using spectroscopic and diffraction-based techniques. In phosphate-free system, uranium oxyhydroxide minerals such as K-compreignacite [K 2 (UO 2 ) 6 O 4 (OH) 6 ·7H 2 O] were formed. Single-pass flow-through (SPFT) and column leaching experiments using synthetic Hanford pore water showed that uranium precipitated as meta-ankoleite during acid weathering was strongly retained in the sediments, with an average release rate of 2.67 × 10 -12 mol g -1 s -1 . In the absence of phosphate, uranium release was controlled by dissolution of uranium oxyhydroxide (compreignacite-type) mineral with a release rate of 1.05-2.42 × 10 -10 mol g -1 s -1 . The uranium mineralogy and release rates determined for both systems in this study support the development of accurate U-release models for the prediction of contaminant transport. These results suggest that phosphate minerals may be a good candidate for uranium remediation approaches at contaminated sites.

Uranium Release from Acidic Weathered Hanford Sediments: Single-Pass Flow-Through and Column Experiments

Energy Technology Data Exchange (ETDEWEB)

Wang, Guohui [Pacific Northwest National Laboratory, Richland, Washington 99354, United States; Um, Wooyong [Pacific Northwest National Laboratory, Richland, Washington 99354, United States; Pohang University of Science and Technology (POSTECH), Pohang, South Korea; Wang, Zheming [Pacific Northwest National Laboratory, Richland, Washington 99354, United States; Reinoso-Maset, Estela [Sierra; Washton, Nancy M. [Pacific Northwest National Laboratory, Richland, Washington 99354, United States; Mueller, Karl T. [Pacific Northwest National Laboratory, Richland, Washington 99354, United States; Perdrial, Nicolas [Department; Department; O’Day, Peggy A. [Sierra; Chorover, Jon [Department

2017-09-21

The reaction of acidic radioactive waste with sediments can induce mineral transformation reactions that, in turn, control contaminant fate. Here, sediment weathering by synthetic uranium-containing acid solutions was investigated using bench-scale experiments to simulate waste disposal conditions at Hanford’s cribs, USA. During acid weathering, the presence of phosphate exerted a strong influence over uranium mineralogy and a rapidly precipitated, crystalline uranium phosphate phase (meta-ankoleite [K(UO2)(PO4)·3H2O]) was identified using spectroscopic and diffraction-based techniques. In phosphate-free system, uranium oxyhydroxide minerals such as K-compreignacite [K2(UO2)6O4(OH)6·7H2O] were formed. Single-pass flow-through (SPFT) and column leaching experiments using synthetic Hanford pore water showed that uranium precipitated as meta-ankoleite during acid weathering was strongly retained in the sediments, with an average release rate of 2.67E-12 mol g-1 s-1. In the absence of phosphate, uranium release was controlled by dissolution of uranium oxyhydroxide (compreignacite-type) mineral with a release rate of 1.05-2.42E-10 mol g-1 s-1. The uranium mineralogy and release rates determined for both systems in this study support the development of accurate U-release models for prediction of contaminant transport. These results suggest that phosphate minerals may be a good candidate for uranium remediation approaches at contaminated sites.

Single-pass continuous-flow leach test of PNL 76-68 glass: some selected Bead Leach I results

International Nuclear Information System (INIS)

Coles, D.G.

1981-01-01

A single-pass continuous-flow leach test of PNL 76-68 glass beads (7 mm dia) was concluded after 420 days of uninterrupted operation. Variables included in the experimental matrix were flow-rate, leachant composition, and temperature. Analysis was conducted on all leachate samples for 237 Np and 239 Pu as well as a number of nonradioactive elements. Results indicated that flow-rate and leachant systematically affected the leach rate, but only slightly. Temperature effects were significant. Plutonium leach rate was lower at higher temperature suggesting that Pu sorption onto the beads was enhanced at the higher temperature. The range of leach rates for all analyzed elements (except Pu), at both temperatures, at all three flow rates, and with all three leachant compositions varied over only three orders of magnitude. The range of variables used in this experiment covered those expected in many proposed repository environments. The preliminary interpretation of the results aPPh 3 also reacted with Mn 2 (CO) 10 and Cp 2 Mo 2 (CO) 6 to give a variety of products at room temperature. A radical mechanism was suggested

Standard practice for measurement of the glass dissolution rate using the single-pass flow-through test method

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice describes a single-pass flow-through (SPFT) test method that can be used to measure the dissolution rate of a homogeneous silicate glass, including nuclear waste glasses, in various test solutions at temperatures less than 100°C. Tests may be conducted under conditions in which the effects from dissolved species on the dissolution rate are minimized to measure the forward dissolution rate at specific values of temperature and pH, or to measure the dependence of the dissolution rate on the concentrations of various solute species. 1.2 Tests are conducted by pumping solutions in either a continuous or pulsed flow mode through a reaction cell that contains the test specimen. Tests must be conducted at several solution flow rates to evaluate the effect of the flow rate on the glass dissolution rate. 1.3 This practice excludes static test methods in which flow is simulated by manually removing solution from the reaction cell and replacing it with fresh solution. 1.4 Tests may be conducted wit...

The integral fast reactor

International Nuclear Information System (INIS)

Till, C.E.

1987-01-01

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

Results of assembly test of HTTR reactor internals

International Nuclear Information System (INIS)

Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

1996-01-01

The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

TPG bus passes

CERN Multimedia

Staff Association

2013-01-01

The CERN Staff Association will stop selling TPG bus passes. All active and retired members of the CERN personnel will be able to purchase Unireso bus passes from the CERN Hostel - Building 39 (Meyrin site) from 1st February 2013. For more information: https://cds.cern.ch/journal/CERNBulletin/2013/04/Announcements/1505279?ln=en

Attenuation of Reactor Gamma Radiation and Fast Neutrons Through Large Single-Crystal Materials

International Nuclear Information System (INIS)

Adib, M.

2009-01-01

A generalized formula is given which, for neutron energies in the range 10-4< E< 10 eV and gamma rays with average energy 2 MeV , permits calculation of the transmission properties of several single crystal materials important for neutron scattering instrumentation. A computer program Filter was developed which permits the calculation of attenuation of gamma radiation, nuclear capture, thermal diffuse and Bragg-scattering cross-sections as a function of materials constants, temperature and neutron energy. The applicability of the deduced formula along with the code checked from the obtained agreement between the calculated and experimental neutron transmission through various single-crystals A feasibility study for use of Si, Ge, Pb, Bi and sapphire is detailed in terms of optimum crystal thickness, mosaic spread and cutting plane for efficient transmission of thermal reactor neutrons and for rejection of the accompanying fast neutrons and gamma rays.

Neutron transmission and reflection at a copper single crystal

Energy Technology Data Exchange (ETDEWEB)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Fayek, M.; Habib, N. (Atomic Energy Establishment, Cairo (Egypt). Reactor and Neutron Physics Dept.); Wahba, M. (Ain Shams Univ., Cairo (Egypt). Dept. of Engineering Physics and Mathematics)

1991-06-01

Neutron transmission and reflection at a copper single crystal cut along the (111) plane were studied with the fixed-scattering-angle spectrometer installed at the ET-RR-1 reactor. The transmission was measured for neutron wavelengths between 0.15 and 0.46 nm and various orientations of the (111) plane with respect to the incident beam. When used as a neutron band pass filter, the crystal is optimally oriented when the neutron beam is incident parallel to the (111) direction. The reflectivity was measured for the (111) plane at 45deg with respect to the incident beam. The results were found to be in reasonable agreement with a value predicted for the reflected intensity at an imperfect crystal with finite absorption. (orig.).

Neutron transmission and reflection at a copper single crystal

International Nuclear Information System (INIS)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Fayek, M.; Habib, N.; Wahba, M.

1991-01-01

Neutron transmission and reflection at a copper single crystal cut along the (111) plane were studied with the fixed-scattering-angle spectrometer installed at the ET-RR-1 reactor. The transmission was measured for neutron wavelengths between 0.15 and 0.46 nm and various orientations of the (111) plane with respect to the incident beam. When used as a neutron band pass filter, the crystal is optimally oriented when the neutron beam is incident parallel to the [111] direction. The reflectivity was measured for the (111) plane at 45deg with respect to the incident beam. The results were found to be in reasonable agreement with a value predicted for the reflected intensity at an imperfect crystal with finite absorption. (orig.) [de

The instrumentation of fast reactor

International Nuclear Information System (INIS)

Endo, Akira

2003-03-01

The author has been engaged in the development of fast reactors over the last 30 years with both an involvement with the early technology development on the experimental breeder reactor Joyo, and latterly continuing this work on the prototype breeder reactor, Monju. In order to pass on this experience to younger engineers this paper is produced to outline this experience in the sincere hope that the information given will be utilised in future educational training material. The paper discusses the wide diversity on the associated instrument technology which the fast breeder reactor requires. The first chapter outlines the fast reactor system, followed by discussions on reactor instrumentation, measurement principles, temperature dependencies, and verification response characteristics from various viewpoints, are discussed in chapters two and three. The important issues of failed fuel location detection, and sodium leak detection from steam generators are discussed in chapters 4 and 5 respectively. Appended to this report is an explanation on the methods of measuring response characteristics on instrumentation systems using error analysis, random signal theory and measuring method of response characteristic by AR (autoregressive) model on which it appears is becoming an indispensable problem for persons involved with this technology in the future. (author)

Scram device for gas-cooled reactor

International Nuclear Information System (INIS)

Murakami, Atsushi; Takahashi, Suehiro.

1989-01-01

A scram device for gas-cooled reactors has a hopper disposed below a stand pipe standing upright passing through a reactor container and electromagnets disposed therein. It further comprises neutron absorbing steel balls maintained between the electromagnets and the hopper upon energization of the electromagnets. Upon emergency reactor shutdown, energization for the electromagnets is interrupted to drop the neutron absorption stainless steel balls into the reactor core. It is an object of the present invention to keep the mechanical strength of the electromagnets in a high temperature gas atmosphere and not to reduce the insulation performance. That is, coils for the electromagnets are constituted with a small oxide-insulated metal sheath cable (MI cable). As the feature of the MI cable, it can maintain the mechanical strength even when exposed to high temperature gas coolant and the insulation performance thereof does not reduce by virture of its gas sealing property. Accordingly, a scram device of stable reliability can be obtained. (K.M.)

Development of a 10-decade single-mode reactor flux monitoring system

International Nuclear Information System (INIS)

Valentine, K.H.; Shepard, R.L.; Falter, K.G.; Reese, W.B.

1988-01-01

Conventional wide-range neutron channels employ three optional modes to monitor the required flux range from source levels to full power (typically 10 or more decades). Difficult calibrations are necessary to provide a continuous output signal when such a system switches from counting mode in the source range to mean-square voltage mode in the midrange to dc current mode in the power range. In an ORNL proof-of-principle test, a method of extended range counting was implemented with a fission counter and conventional wide-band pulse processing electronics to provide a single-mode, monotonically increasing signal that spanned /approximately 10/ decades of neutron flux. Ongoing work includes design, fabrication, and testing of a comlpete neutron flux monitoring system suitable for advanced liquid metal reactor designs. 6 refs., 4 figs

Loop type LMFBR reactor

International Nuclear Information System (INIS)

Ito, Hiroyuki

1989-01-01

In conventional FBR type reactors, primary coolants at high temperature uprise at a great flow rate and, due to the dynamic pressure thereof, the free surface is raised or sodium is partially jetted upwardly and then fallen again. Then, a wave killing plate comprising a buffer plate and a deflection plate is disposed to the liquid surface of coolants. Most of primary sodium uprising from the reactor core along the side of the upper mechanism during operation collide against the buffer plate of the wave killing plate to moderate the dynamic pressure and, further, disperse radially of the reactor vessel. On the other hand, primary sodium passing through flowing apertures collides against the deflection plate opposed to the flowing apertures to moderate the dynamic pressure, by which the force of raising the free surface is reduced. Thus, uprising and waving of the free surface can effectively be suppressed to reduce the incorporation of cover gases into the primary sodium, so that it is possible to prevent in injury of the recycling pump, abrupt increase of the reactor core reactivity and reduction of the heat efficiency of intermediate heat exchangers. (N.H.)

Reactor cooling apparatus

International Nuclear Information System (INIS)

Ogura, Kenji.

1983-01-01

Purpose: To increase natural convection flowrate in the reactor core upon interruption of a recycling pump by remarkably decreasing the flow resistance. Constitution: By-pass lines are disposed to a recycling pump in a primary coolant system and a second recycling pump in a secondary coolant system respectively, and a check valve and an isolation valve are attached to each of them. Each of the isolation valves is closed during normal operation and automatically opened when the number of rotation for each of the recycling pumps goes lower than a predetermined value. This can significantly decrease the flow resistance in the primary and secondary coolant systems upon interruption of the recycling pumps due to the entire loss of AC power source or the like to thereby increase the natural convection flowrate in the reactor core. (Sekiya, K.)

A structure for the protection of nuclear-reactor pressurized-vessels against rupture

International Nuclear Information System (INIS)

Marcellin, J.-P.; Aubert, Gilles

1974-01-01

Description is given of a structure for the protection of nuclear-reactor pressurized-vessels against rupture. Said structure comprises a pre-stressed concrete tank adapted to surround the tank side-wall and bottom, said tank being higher than said vessel, said tank being provided with ports for passing cooling fluid ducts therethrough, and a crown adapted to rest along the periphery of the reactor-cover and made integral therewith. This can be applied to reactors of the PWR type [fr

WebPASS Explorer (HR Personnel Management)

Data.gov (United States)

US Agency for International Development — WebPass Explorer (WebPASS Framework): USAID is partnering with DoS in the implementation of their WebPass Post Personnel (PS) Module. WebPassPS does not replace...

Steam generator of FBR type reactor

International Nuclear Information System (INIS)

Hashiguchi, Ko.

1992-01-01

Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

Method of shielding a liquid-metal-cooled reactor

International Nuclear Information System (INIS)

Sayre, R.K.

1978-01-01

The primary heat transport system of a nuclear reactor - particularly for a liquid-metal-cooled fast-breeder reactor - is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of a the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system

Single phase and two phase erosion corrosion in broilers of gas-cooled reactors

International Nuclear Information System (INIS)

Harrison, G.S.; Fountain, M.J.

1988-01-01

Erosion-corrosion is a phenomenon causing metal wastage in a variety of locations in water and water-steam circuits throughout the power generation industry. Erosion-corrosion can occur in a number of regions of the once-through boiler designs used in the later Magnox and AGR type of gas cooled nuclear reactor. This paper will consider two cases of erosion-corrosion damage (single and two phase) in once through boilers of gas cooled reactors and will describe the solutions that have been developed. The single phase problem is associated with erosion-corrosion damage of mild steel downstream of a boiler inlet flow control orifice. With metal loss rates of up to 1 mm/year at 150 deg. C and pH in the range 9.0-9.4 it was found that 5 μg/kg oxygen was sufficient to reduce erosion-corrosion rates to less than 0.02 mm/year. A combined oxygen-ammonia-hydrazine feedwater regime was developed and validated to eliminate oxygen carryover and hence give protection from stress corrosion in the austenitic section of the AGR once through boiler whilst still providing erosion-corrosion control. Two phase erosion-corrosion tube failures have occurred in the evaporator of the mild steel once through boilers of the later Magnox reactors operating at pressures in the range 35-40 bar. Rig studies have shown that amines dosed in the feedwater can provide a significant reduction in metal loss rates and a tube lifetime assessment technique has been developed to predict potential tube failure profiles in a fully operational boiler. The solutions identified for both problems have been successfully implemented and the experience obtained following implementation including any problems or other benefits arising from the introduction of the new regimes will be presented. Methods for monitoring and evaluating the efficiency of the solutions have been developed and the results from these exercises will also be discussed. Consideration will also be given to the similarities in the metal loss

Single phase and two phase erosion corrosion in broilers of gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Harrison, G S; Fountain, M J [Operational Engineering Division (Northern Area), Central Electricity Generating Board, Manchester (United Kingdom)

1988-07-01

Erosion-corrosion is a phenomenon causing metal wastage in a variety of locations in water and water-steam circuits throughout the power generation industry. Erosion-corrosion can occur in a number of regions of the once-through boiler designs used in the later Magnox and AGR type of gas cooled nuclear reactor. This paper will consider two cases of erosion-corrosion damage (single and two phase) in once through boilers of gas cooled reactors and will describe the solutions that have been developed. The single phase problem is associated with erosion-corrosion damage of mild steel downstream of a boiler inlet flow control orifice. With metal loss rates of up to 1 mm/year at 150 deg. C and pH in the range 9.0-9.4 it was found that 5 {mu}g/kg oxygen was sufficient to reduce erosion-corrosion rates to less than 0.02 mm/year. A combined oxygen-ammonia-hydrazine feedwater regime was developed and validated to eliminate oxygen carryover and hence give protection from stress corrosion in the austenitic section of the AGR once through boiler whilst still providing erosion-corrosion control. Two phase erosion-corrosion tube failures have occurred in the evaporator of the mild steel once through boilers of the later Magnox reactors operating at pressures in the range 35-40 bar. Rig studies have shown that amines dosed in the feedwater can provide a significant reduction in metal loss rates and a tube lifetime assessment technique has been developed to predict potential tube failure profiles in a fully operational boiler. The solutions identified for both problems have been successfully implemented and the experience obtained following implementation including any problems or other benefits arising from the introduction of the new regimes will be presented. Methods for monitoring and evaluating the efficiency of the solutions have been developed and the results from these exercises will also be discussed. Consideration will also be given to the similarities in the metal loss

Asymptotic estimation of reactor fueling optimal strategy

International Nuclear Information System (INIS)

Simonov, V.D.

1985-01-01

The problem of improving the technical-economic factors of operating. and designed nuclear power plant blocks by developino. internal fuel cycle strategy (reactor fueling regime optimization), taking into account energy system structural peculiarities altogether, is considered. It is shown, that in search of asymptotic solutions of reactor fueling planning tasks the model of fuel energy potential (FEP) is the most ssuitable and effective. FEP represents energy which may be produced from the fuel in a reactor with real dimensions and power, but with hypothetical fresh fuel supply, regime, providing smilar burnup of all the fuel, passing through the reactor, and continuous overloading of infinitely small fuel portion under fule power, and infinitely rapid mixing of fuel in the reactor core volume. Reactor fuel run with such a standard fuel cycle may serve as FEP quantitative measure. Assessment results of optimal WWER-440 reactor fresh fuel supply periodicity are given as an example. The conclusion is drawn that with fuel enrichment x=3.3% the run which is 300 days, is economically justified, taking into account that the cost of one energy unit production is > 3 cop/KW/h

Thermohydraulics of reactors

International Nuclear Information System (INIS)

Delhaye, J.M.

2008-01-01

This scientific and technical handbook about PWR reactors thermohydraulics is the result of many years of teaching in the framework of the CEA-INSTN's atomic engineering training courses, in engineering schools and during continuing training activities. Its main goal is to present in a rigorous and pedagogical way the basic knowledge necessary for the understanding and modeling of single phase and two-phase thermohydraulic phenomena encountered during the design and operation of nuclear reactors. In particular, heat transfers in two-phase flows are presented in a detailed way. Most chapters include some nuclear engineering examples of application of the studied concepts, and some exercises aiming at mastering these concepts. Each example or exercise is accompanied by its detailed solution. Content: - thermohydraulic characteristics of reactors; - design and thermal dimensioning of reactors; - thermal engineering of the fuel element; - two-phase flow configurations in ducts; - recalls about single-phase flow equations; - basic equations for two-phase flows; - modeling of two-phase flows inside ducts; - pressure drops in ducts; - boiling and vapor condensation heat transfers; - two-phase flow instabilities in ducts; - two-phase flow blocking; thermohydraulics of naval propulsion reactors

Disposition of the fluoride fuel and flush salts from the Molten Salt Reactor experiment at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Peretz, F.J.

1996-01-01

The Molten Salt Reactor Experiment (MSRE) is an 8 MW reactor that was operated at Oak Ridge National Laboratory (ORNL) from 1965 through 1969. The reactor used a unique liquid salt fuel, composed of a mixture of LIF, BeF 2 , ZrF 4 , and UF 4 , and operated at temperatures above 600 degrees C. The primary fuel salt circulation system consisted of the reactor vessel, a single fuel salt pump, and a single primary heat exchanger. Heat was transferred from the fuel salt to a coolant salt circuit in the primary heat exchanger. The coolant salt was similar to the fuel salt, except that it contains only LiF (66%) and BeF, (34%). The coolant salt passed from the primary heat exchanger to an air-cooled radiator and a coolant salt pump, and then returned to the primary heat exchanger. Each of the salt loops was provided with drain tanks, located such that the salt could be drained out of either circuit by gravity. A single drain tank was provided for the non-radioactive coolant salt. Two drain tanks were provided for the fuel salt. Since the fuel salt contained radioactive fuel, fission products, and activation products, and since the reactor was designed such that the fuel salt could be drained immediately into the drain tanks in the event of a problem in the fuel salt loop, the fuel salt drain tanks were provided with a system to remove the heat generated by radioactive decay. A third drain tank connected to the fuel salt loop was provided for a batch of flush salt. This batch of salt, similar in composition to the coolant salt, was used to condition the fuel salt loop after it had been exposed to air and to flush the fuel salt loop of residual fuel salt prior to accessing the reactor circuit for maintenance or experimental activities. This report discusses the disposition of the fluoride fuel and flush salt

Chemical reactor for converting a first material into a second material

Science.gov (United States)

Kong, Peter C

2012-10-16

A chemical reactor and method for converting a first material into a second material is disclosed and wherein the chemical reactor is provided with a feed stream of a first material which is to be converted into a second material; and wherein the first material is combusted in the chemical reactor to produce a combustion flame, and a resulting gas; and an electrical arc is provided which is passed through or superimposed upon the combustion flame and the resulting gas to facilitate the production of the second material.

Plenum separator system for pool-type nuclear reactors

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1983-01-01

This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

Removal of tritium from gas-cooled nuclear reactors

International Nuclear Information System (INIS)

Nieder, R.

1976-01-01

Tritium contained in the coolant gas in the primary circuit of a gas cooled nuclear reactor together with further tritium adsorbed on the graphite used as a moderator for the reactor is removed by introducing hydrogen or a hydrogen-containing compound, for example methane or ammonia, into the coolant gas. The addition of the hydrogen or hydrogen-containing compound to the coolant gas causes the adsorbed tritium to be released into the coolant gas and the tritium is then removed from the coolant gas by passing the mixture of coolant gas and hydrogen or hydrogen-containing compound through a gas purification plant before recirculating the coolant gas through the reactor. 14 claims, 1 drawing figure

Instruments for non-destructive evaluation of advanced test reactor inpile tubes

International Nuclear Information System (INIS)

Livingston, R.A.; Beller, L.S.; Edgett, S.M.

1986-01-01

The Advanced Test Reactor is a 250 MW LWR used primarily for irradiation testing of materials contained in inpile tubes that pass through the reactor core. These tubes provided the high pressure and temperature water environment required for the test specimens. The reactor cooling water surrounding the inpile tubes is at much lower pressure and temperature. The structural integrity of the inpile tubes is monitored by routine surveillance to ensure against unplanned reactor shutdowns to replace defective inpile tubes. The improved instruments developed for inpile tube surveillance include a bore profilometer, ultrasonic flaw detetion system and bore diameter gauges. The design and function of these improved instruments is presented

Microstructure evolution of pure copper during a single pass of simple shear extrusion (SSE): role of shear reversal

Energy Technology Data Exchange (ETDEWEB)

Bagherpour, E., E-mail: e.bagherpour@semnan.ac.ir [Faculty of Metallurgical and Materials Engineering, Semnan University, Semnan (Iran, Islamic Republic of); Department of Mechanical Engineering, Doshisha University, Kyotanabe, Kyoto 610–0394 (Japan); Qods, F., E-mail: qods@semnan.ac.ir [Faculty of Metallurgical and Materials Engineering, Semnan University, Semnan (Iran, Islamic Republic of); Ebrahimi, R., E-mail: ebrahimy@shirazu.ac.ir [Department of Materials Science and Engineering, School of Engineering, Shiraz University, Shiraz (Iran, Islamic Republic of); Miyamoto, H., E-mail: hmiyamot@mail.doshisha.ac.jp [Department of Mechanical Engineering, Doshisha University, Kyotanabe, Kyoto 610–0394 (Japan)

2016-06-01

In the present paper the role of shear reversal on microstructure, texture and mechanical properties of pure copper during a single pass of the simple shear extrusion (SSE) process was investigated. For SSE processing an appropriate die with a linear die profile was designed and constructed, which imposes forward shear in the first half and reverse shear in the second half channels. Electron back-scattering diffraction (EBSD), transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) were used to evaluate the microstructure of the deformed samples. The geometrical nature of this process imposes a distribution of strain results in the inhomogeneous microstructure and the hardness throughout the plane perpendicular to the extrusion direction. Strain reversal during the process results in a slight reduction in dislocation density, the hardness and mean disorientation angle of the samples, and an increase in the grain size. After a complete pass of SSE, dislocation density decreased by ~14% if compared to the middle of the process. This suggests that the dislocation annihilation occurred by the reversal of the shear strain. The simple shear textures were formed gradually and the strongest simple shear textures were observed on the middle of the SSE channel. The degree of the simple shear textures decreases with the distance from the middle plane where the shear is reversed, but the simple shear textures are still the major components after exit of the channel. Hardness variation was modeled by contributions from dislocation strengthening and grain boundary strengthening, where dislocation density is approximated by the misorientation angle of LAGBs which are regarded as dislocation cell boundaries. As a result, the hardness can be predicted successfully by the microstructural features, i.e. the low-angle boundaries, the mean misorientation angle and the fraction of high-angle grain boundaries.

Elements for evaluation of fast breeder reactor's potential in Argentina

International Nuclear Information System (INIS)

Gho, C.J.

1985-01-01

Fast Breeder Reactors (FBR) main features are presented in a general form, including their physical principles, the history of their evolution, their relevant technological aspects and the basis for their comparison to other energy sources. This is completed with descriptions of typical reactors and a model of FBR penetration in the Argentine electrical network. It is recommended to form a multidisciplinary board to study which position should be taken with respect to this type of reactors. In the author's opinion a Research activity should be started and gradually increased for passing to Development activities after a short while. (Author) [es

Evaluation of Two Passes Cold Pilgering Property for PLUS7TM Guide Thimble and Instrumentation Tubing

Energy Technology Data Exchange (ETDEWEB)

Park, Min Young; Park, Ki Bum; Kim, In Kyu; Lee, Young Hee; Kahng, Jong Yeol [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

2015-05-15

The thermo-mechanical property of zirconium alloy tube is well known to be influenced by pilgering pass schedule and its tooling; thus the control of its microstructure and mechanical property in the final tube production stage for nuclear fuel applications is a major concern of tube manufacture. To fabricate final tube, the 3 passes pilgering is applied in general by using TREX(Tube Reduced EXtrusion), 63.5mm outer diameter(OD), in KEPCO NF and most of Zr tube manufacturing companies. They are also taking big efforts to reduce pilgering step for the sake of increasing the efficiency of production in the forming stage of tube. The objective of this study is to develop two passes of pilgering schedule from the conventional three passes of pilgering schedule for manufacturing the Guide Thimble and Instrumentation tube conforming to specification, which are newly developing component for the advanced nuclear fuel assembly in KEPCO NF. CSR, hydride orientation, and structural integrity are well conformed to the desired targets so it is expected that both die and mandrel were newly designed for the PLUS7TM guide thimble and instrumentation tube with higher Q factor for two passes of pilgering at 50LC and 25LC pilger machine, instead of three passes of pilgering, are able to be applicable to this design of fuel component. If developed two passes pilgering is applied to current manufacturing process, it would improve not only productivity but also yield rate by reducing 3 steps(pilgering, heat-treatment, pickiling and cleaning) of manufacturing process. But additional tests(including in-pile test) should be performed in order to evaluate integrity in reactor.

"Which pass is better?" Novel approaches to assess passing effectiveness in elite soccer.

Science.gov (United States)

Rein, Robert; Raabe, Dominik; Memmert, Daniel

2017-10-01

Passing behaviour is a key property of successful performance in team sports. Previous investigations however have mainly focused on notational measurements like total passing frequencies which provide little information about what actually constitutes successful passing behaviour. Consequently, this has hampered the transfer of research findings into applied settings. Here we present two novel approaches to assess passing effectiveness in elite soccer by evaluating their effects on majority situations and space control in front of the goal. Majority situations are assessed by calculating the number of defenders between the ball carrier and the goal. Control of space is estimated using Voronoi-diagrams based on the player's positions on the pitch. Both methods were applied to position data from 103 German First division games from the 2011/2012, 2012/2013 and 2014/2015 seasons using a big data approach. The results show that both measures are significantly related to successful game play with respect to the number of goals scored and to the probability of winning a game. The results further show that on average passes from the mid-field into the attacking area are most effective. The presented passing efficiency measures thereby offer new opportunities for future applications in soccer and other sports disciplines whilst maintaining practical relevance with respect to tactical training regimes or game performances analysis. Copyright © 2017 Elsevier B.V. All rights reserved.

Single-phase AutoReClosure ARC failure on 400 kV combinedcable/overhead line with permanently connected shunt reactor

DEFF Research Database (Denmark)

Bak, Claus Leth; Søgaard, Kim

2008-01-01

consisting of overhead lines, crossbonded cable sections and shunt reactor has been created in PSCAD/EMTDC and verified against measurements with good results. Main focus has been put on the likelihood of having a successful single-phase autoreclosure ARC in such a combined cable/OHL line....

Direct measurement of first-pass ileal clearance of a bile acid in humans

International Nuclear Information System (INIS)

Galatola, G.; Jazrawi, R.P.; Bridges, C.; Joseph, A.E.; Northfield, T.C.

1991-01-01

The purpose of this study was to develop and validate a method of directly measuring ileal bile acid absorption efficiency during a single enterohepatic cycle (first-pass ileal clearance). This has become feasible for the first time because of the availability of the synthetic gamma-labeled bile acid 75Selena-homocholic acid-taurine (75SeHCAT). Together with the corresponding natural bile acid cholic acid-taurine (labeled with 14C), SeHCAT was infused distal to an occluding balloon situated beyond the ampulla of Vater in six healthy subjects. Completion of a single enterohepatic cycle was assessed by obtaining a plateau for 75SeHCAT activity proximal to the occluding balloon, which prevented further cycles. Unabsorbed 75SeHCAT was collected after total gut washout, which was administered distal to the occluding balloon. 75SeHCAT activity in the rectal effluent measured by gamma counter was compared with that of absorbed 75SeHCAT level measured by gamma camera and was used to calculate first-pass ileal clearance. This was very efficient (mean value, 96%) and showed very little variation in the six subjects studied (range, 95%-97%). A parallel time-activity course in hepatic bile for 14C and 75Se during a single enterohepatic cycle, together with a ratio of unity for 14C/75Se in samples obtained at different time intervals, suggests that 75SeHCAT is handled by the ileum like the natural bile acid cholic acid-taurine. Extrapolation of 75SeHCAT first-pass ileal clearance to that of the natural bile acid therefore seems justifiable. In a subsidiary experiment, ileal absorption efficiency per day for 75SeHCAT was also measured by scanning the gallbladder area on 5 successive days after the measurement of first-pass ileal clearance. In contrast with absorption efficiency per cycle, absorption efficiency per day varied widely (49%-86%)

Regulation for installation and operation of reactor

International Nuclear Information System (INIS)

1977-01-01

Concerning the description of an application for the approval of installation of a reactor, stipulated in Article 23 paragraph 2 of the Law for Regulation of Nuclear Source Materials, Nuclear Fuel Materials and Reactors (hereinafter referred to as the Law), the following items must be written. Namely, the heat output of the reactor in Article 23 paragraph 2 item 3 of the Law, the position, structure and facilities of the reactor facilities, described according to the stipulated classifications, the work plan, nuclear fuel materials employed, and the disposal of spent fuel. Concerning an application for the approval of a reactor installed aboard a foreign ship, stipulations are made separately. Description of an application for the approval of change of the heat output of a reactor and others should include the stipulated items. When it is wished to undergo inspection of the construction and performance of reactor facilities, an application for that end including the required items should be filed. Various safety measures preventing personnel from being exposed to radiation should be taken. When a foreign atomic-powered ship tries to enter a Japanese port, the stipulated necessary informations should be reported 60 days before such ship actually enters the Japanese port. A chief technician of reactors should take and pass the official examination. (Rikitake, Y.)

A recycling molecular beam reactor

International Nuclear Information System (INIS)

Prada-Silva, G.; Haller, G.L.; Fenn, J.B.

1974-01-01

In a Recycling Molecular Beam Reactor, RMBR, a beam of reactant gas molecules is formed from a supersonic free jet. After collision with a target the molecules pass through the vacuum pumps and are returned to the nozzle source. Continuous recycling permits the integration of very small reaction probabilities into measurable conversions which can be analyzed by gas chromatography. Some preliminary experiments have been carried out on the isomerization of cyclopropane

Testing of plain and fibrous concrete single cavity prestressed concrete reactor vessel models

International Nuclear Information System (INIS)

Oland, C.B.

1985-01-01

Two single-cavity prestressed concrete reactor vessel (PCRV) models were fabricated and tested to failure to demonstrate the structural response and ultimate pressure capacity of models cast from high-strength concretes. Concretes with design compressive strengths in excess of 70 MPa (10,000 psi) were developed for this investigation. One model was cast from plain concrete and failed in shear at the head region. The second model was cast from fiber reinforced concrete and failed by rupturing the circumferential prestressing at the sidewall of the structure. The tests also demonstrated the capabilities of the liner system to maintain a leak-tight pressure boundary. 3 refs., 4 figs

Optimization of TRU burnup in modular helium reactor

International Nuclear Information System (INIS)

Yonghee, Kim; Venneri, F.

2007-01-01

An optimization study of a single-pass TRU (transuranic) deep-burn (DB) has been performed for a block-type MHR (Modular Helium Reactor) proposed by General Atomics. Assuming a future equilibrium scenario of advanced LWRs, a high-burnup TRU vector is considered: 50 GWD/MTU and 5-year cooling. For 3-D equilibrium cores, the performance analysis is done by using a continuous energy Monte Carlo depletion code MCCARD. The core optimization is performed from the viewpoints of the core configuration, fuel management, TRISO fuel specification, and neutron spectrum. With regard to core configuration, two annular cores are investigated in terms of the neutron economy. A conventional radial shuffling scheme of fuel blocks is compared with an axial block shuffling strategy in terms of the fuel burnup and core power distributions. The impact of the kernel size of TRISO fuel is evaluated and a diluted kernel, instead of a conventional concentrated kernel, is introduced to maximize the TRU burnup by reducing the self-shielding effects of TRISO fuels. A higher graphite density is evaluated in terms of the fuel burnup. In addition, it is shown that the core power distribution can be effectively controlled by zoning of the packing fraction of TRISO fuels. We also have shown that a long-cycle DB-MHR core can be designed by using a small batch size for fuel reloading, at the expense of a marginal decrease of the TRU discharge burnup. Depending on the fuel management scheme, fuel specifications, and core parameters, the TRU burnup in an optimized DB-MHR core is over 60% in a single-pass irradiation campaign. (authors)

Wave-optics simulation of the double-pass beam propagation in modulating retro-reflector FSO systems using a corner cube reflector.

Science.gov (United States)

Yang, Guowei; You, Shengzui; Bi, Meihua; Fan, Bing; Lu, Yang; Zhou, Xuefang; Li, Jing; Geng, Hujun; Wang, Tianshu

2017-09-10

Free-space optical (FSO) communication utilizing a modulating retro-reflector (MRR) is an innovative way to convey information between the traditional optical transceiver and the semi-passive MRR unit that reflects optical signals. The reflected signals experience turbulence-induced fading in the double-pass channel, which is very different from that in the traditional single-pass FSO channel. In this paper, we consider the corner cube reflector (CCR) as the retro-reflective device in the MRR. A general geometrical model of the CCR is established based on the ray tracing method to describe the ray trajectory inside the CCR. This ray tracing model could treat the general case that the optical beam is obliquely incident on the hypotenuse surface of the CCR with the dihedral angle error and surface nonflatness. Then, we integrate this general CCR model into the wave-optics (WO) simulation to construct the double-pass beam propagation simulation. This double-pass simulation contains the forward propagation from the transceiver to the MRR through the atmosphere, the retro-reflection of the CCR, and the backward propagation from the MRR to the transceiver, which can be realized by a single-pass WO simulation, the ray tracing CCR model, and another single-pass WO simulation, respectively. To verify the proposed CCR model and double-pass WO simulation, the effective reflection area, the incremental phase, and the reflected beam spot on the transceiver plane of the CCR are analyzed, and the numerical results are in agreement with the previously published results. Finally, we use the double-pass WO simulation to investigate the double-pass channel in the MRR FSO systems. The histograms of the turbulence-induced fading in the forward and backward channels are obtained from the simulation data and are fitted by gamma-gamma (ΓΓ) distributions. As the two opposite channels are highly correlated, we model the double-pass channel fading by the product of two correlated �

Initiating Events for Multi-Reactor Plant Sites

Energy Technology Data Exchange (ETDEWEB)

Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-09-01

Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Impact of one-to-one tutoring on fundamentals of laparoscopic surgery (FLS) passing rate in a single center experience outside the United States: a randomized controlled trial.

Science.gov (United States)

Gheza, Federico; Raimondi, Paolo; Solaini, Leonardo; Coccolini, Federico; Baiocchi, Gian Luca; Portolani, Nazario; Tiberio, Guido Alberto Massimo

2018-04-11

Outside the US, FLS certification is not required and its teaching methods are not well standardized. Even if the FLS was designed as "stand alone" training system, most of Academic Institution offer support to residents during training. We present the first systematic application of FLS in Italy. Our aim was to evaluate the role of mentoring/coaching on FLS training in terms of the passing rate and global performance in the search for resource optimization. Sixty residents in general surgery, obstetrics & gynecology, and urology were selected to be enrolled in a randomized controlled trial, practicing FLS with the goal of passing a simulated final exam. The control group practiced exclusively with video material from SAGES, whereas the interventional group was supported by a mentor. Forty-six subjects met the requirements and completed the trial. For the other 14 subjects no results are available for comparison. One subject for each group failed the exam, resulting in a passing rate of 95.7%, with no obvious differences between groups. Subgroup analysis did not reveal any difference between the groups for FLS tasks. We confirm that methods other than video instruction and deliberate FLS practice are not essential to pass the final exam. Based on these results, we suggest the introduction of the FLS system even where a trained tutor is not available. This trial is the first single institution application of the FLS in Italy and one of the few experiences outside the US. Trial Number: NCT02486575 ( https://www.clinicaltrials.gov ).

Effect of Co-60 single escape peak on detection of Cs-137 in analysis of radionuclide from research reactor

International Nuclear Information System (INIS)

Kim, M. S.; Park, S. J.

2006-01-01

The effect of the single escape peak of 1173 keV gamma-rays from Co-60 on the detection of Cs-137 activity is analyzed. The single escape peak of 1173 keV gamma-rays from Co-60 is located at the 662 keV, which is very close to the energy of gamma-rays from Cs-137. This single escape peak may be mistaken for the gamma-ray peak from Cs-137 activity in the case of large area of 1173 keV peak. The detection of Cs-137 is very important to the judgment of the contamination or the leakage of the material containing the fission product like reactor pool water and in the several experiments for reactor physics such as burn-up estimation. In this work, the areas of the single escape peak of the 1173 keV gamma-rays from Co-60 are measured with several full energy peak areas by using the HPGe detector. The critical limit by which we can decide whether the net count of 662 keV peak due to Co-60 would be significant or not is deduced. For this detection system, when the area of full energy peak is larger than 4.5 million, the single escape peak of 1173 keV gamma-rays from Co-60 can be regarded as the single significant peak. Therefore, it is confirmed that the detection of the Cs-137 activity is affected by the Co-60 in this case. Conservatively, for this detection system, it is recommended that the area of 1173 keV peak of Co-60 would be less than 2 million for neglecting the effect of Co-60. (authors)

Thermosyphoning in the CANDU reactor

International Nuclear Information System (INIS)

Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

1984-01-01

Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

Iron crud supply device to reactor coolant

International Nuclear Information System (INIS)

Baba, Takao.

1993-01-01

In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

Developing a framework for a sustainable research reactor program in the UAE

Energy Technology Data Exchange (ETDEWEB)

Almarri, Khalid [Senate Member and Representative of Students of PhD Program in Project Management, Ajman (United Arab Emirates)

2013-07-01

In 2009, the UAE passed a milestone of preparations for the involvement in the nuclear area by awarding its first nuclear power plant. And to maximise its nuclear benefits, the country needs to develop a research reactor program. This paper's objectives are: a) Selecting and analysing model case studies from other nations to establish the success factors of research reactors in the UAE, and the preparedness of local industries to maximise the potential of the reactors. b) Establishing how the UAE's research reactors will contribute to the coalition and sharing of good practices with other countries as promoted by IAEA.

Mirror fusion reactors

International Nuclear Information System (INIS)

Carlson, G.A.; Moir, R.W.

1978-01-01

We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

Mirror fusion reactors

International Nuclear Information System (INIS)

Anon.

1978-01-01

Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

Feasibility study of multi-pass respiratory-gated helical tomotherapy of a moving target via binary MLC closure

Science.gov (United States)

Kim, Bryan; Chen, Jeff; Kron, Tomas; Battista, Jerry

2010-11-01

Gated radiotherapy of lung lesions is particularly complex for helical tomotherapy, due to the simultaneous motions of its three subsystems (gantry, couch and collimator). We propose a new way to implement gating for helical tomotherapy, namely multi-pass respiratory gating. In this method, gating is achieved by delivering only the beam projections that occur within a respiratory gating window, while blocking the rest of the beam projections by fully closing all collimator leaves. Due to the continuous couch motion, the planned beam projections must be delivered over multiple passes of radiation deliveries. After each pass, the patient couch is reset to its starting position, and the treatment recommences at a different phase of tumour motion to 'fill in' the previously blocked beam projections. The gating process may be repeated until the plan dose is delivered (full gating), or halted after a certain number of passes, with the entire remaining dose delivered in a final pass without gating (partial gating). The feasibility of the full gating approach was first tested for sinusoidal target motion, through experimental measurements with film and computer simulation. The optimal gating parameters for full and partial gating methods were then determined for various fractionation schemes through computer simulation, using a patient respiratory waveform. For sinusoidal motion, the PTV dose deviations of -29 to 5% observed without gating were reduced to range from -1 to 3% for a single fraction, with a 4 pass full gating. For a patient waveform, partial gating required fewer passes than full gating for all fractionation schemes. For a single fraction, the maximum allowed residual motion was only 4 mm, requiring large numbers of passes for both full (12) and partial (7 + 1) gating methods. The number of required passes decreased significantly for 3 and 30 fractions, allowing residual motion up to 7 mm. Overall, the multi-pass gating technique was shown to be a promising

Feasibility study of multi-pass respiratory-gated helical tomotherapy of a moving target via binary MLC closure

Energy Technology Data Exchange (ETDEWEB)

Kim, Bryan; Chen, Jeff; Battista, Jerry [London Regional Cancer Program, London Health Sciences Centre, London, ON (Canada); Kron, Tomas, E-mail: bryan.kim@lhsc.on.c [Peter MacCallum Cancer Center, Melbourne (Australia)

2010-11-21

Gated radiotherapy of lung lesions is particularly complex for helical tomotherapy, due to the simultaneous motions of its three subsystems (gantry, couch and collimator). We propose a new way to implement gating for helical tomotherapy, namely multi-pass respiratory gating. In this method, gating is achieved by delivering only the beam projections that occur within a respiratory gating window, while blocking the rest of the beam projections by fully closing all collimator leaves. Due to the continuous couch motion, the planned beam projections must be delivered over multiple passes of radiation deliveries. After each pass, the patient couch is reset to its starting position, and the treatment recommences at a different phase of tumour motion to 'fill in' the previously blocked beam projections. The gating process may be repeated until the plan dose is delivered (full gating), or halted after a certain number of passes, with the entire remaining dose delivered in a final pass without gating (partial gating). The feasibility of the full gating approach was first tested for sinusoidal target motion, through experimental measurements with film and computer simulation. The optimal gating parameters for full and partial gating methods were then determined for various fractionation schemes through computer simulation, using a patient respiratory waveform. For sinusoidal motion, the PTV dose deviations of -29 to 5% observed without gating were reduced to range from -1 to 3% for a single fraction, with a 4 pass full gating. For a patient waveform, partial gating required fewer passes than full gating for all fractionation schemes. For a single fraction, the maximum allowed residual motion was only 4 mm, requiring large numbers of passes for both full (12) and partial (7 + 1) gating methods. The number of required passes decreased significantly for 3 and 30 fractions, allowing residual motion up to 7 mm. Overall, the multi-pass gating technique was shown to be a

Evaluation of the single-pass flow-through test to support a low-activity waste specification

International Nuclear Information System (INIS)

McGrail, B.P.; Peeler, D.K.

1995-09-01

A series of single-pass flow-through (SPFT) tests was performed on five reference low-activity waste glasses and a reference glass from the National Institute of Standards and Technology to support a product specification for low-activity waste (LAW) forms. The results showed that the SPFT test provides a means to quantitatively distinguish among LAW glass forms in terms of their forward reaction rate at a given temperature and solution pH. Two of the test glasses were also subjected to SPFT testing at Argonne National Laboratory (ANL). Forward reaction rate constants calculated from the ANL test data were 100 to over 1,000 times larger than the values obtained from the SPFT tests conducted at PNL. An analysis of the ANL results showed that they were inconsistent with independent measurements done on glasses of similar composition, the known pH-dependence of the forward rate, and with the results from low surface-area-to-volume, short duration product consistency tests. Because the data set obtained from the SPFT tests done at PNL was consistent with each of these same factors, a detailed examination of the test procedures used at both laboratories was performed to determine the cause(s) of the discrepancy. The omission of background subtraction in the data analysis procedure and the short-duration (on the order of hours) of the ANL tests are factors that may have significantly affected the calculated rates

Nuclear reactor internals arrangement

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

Modeling Geometric Arrangements of TiO2-Based Catalyst Substrates and Isotropic Light Sources to Enhance the Efficiency of a Photocatalystic Oxidation (PCO) Reactor

Science.gov (United States)

Richards, Jeffrey T.; Levine, Lanfang H.; Husk, Geoffrey K.

2011-01-01

The closed confined environments of the ISS, as well as in future spacecraft for exploration beyond LEO, provide many challenges to crew health. One such challenge is the availability of a robust, energy efficient, and re-generable air revitalization system that controls trace volatile organic contaminants (VOCs) to levels below a specified spacecraft maximum allowable concentration (SMAC). Photocatalytic oxidation (PCO), which is capable of mineralizing VOCs at room temperature and of accommodating a high volumetric flow, is being evaluated as an alternative trace contaminant control technology. In an architecture of a combined air and water management system, placing a PCO unit before a condensing heat exchanger for humidity control will greatly reduce the organic load into the humidity condensate loop ofthe water processing assembly (WPA) thereby enhancing the life cycle economics ofthe WPA. This targeted application dictates a single pass efficiency of greater than 90% for polar VOCs. Although this target was met in laboratory bench-scaled reactors, no commercial or SBIR-developed prototype PCO units examined to date have achieved this goal. Furthermore, the formation of partial oxidation products (e.g., acetaldehyde) was not eliminated. It is known that single pass efficiency and partial oxidation are strongly dependent upon the contact time and catalyst illumination, hence the requirement for an efficient reactor design. The objective of this study is to maximize the apparent contact time and illuminated catalyst surface area at a given reactor volume and volumetric flow. In this study, a Ti02-based photocatalyst is assumed to be immobilized on porous substrate panels and illumination derived from linear isotropic light sources. Mathematical modeling using computational fluid dynamics (CFD) analyses were performed to investigate the effect of: 1) the geometry and configuration of catalyst-coated substrate panels, 2) porosity of the supporting substrate, and 3

Nuclear reactor power control device

International Nuclear Information System (INIS)

Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

1987-01-01

Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

Three-Input Single-Output Voltage-Mode Multifunction Filter with Electronic Controllability Based on Single Commercially Available IC

Directory of Open Access Journals (Sweden)

Supachai Klungtong

2017-01-01

Full Text Available This paper presents a second-order voltage-mode filter with three inputs and single-output voltage using single commercially available IC, one resistor, and two capacitors. The used commercially available IC, called LT1228, is manufactured by Linear Technology Corporation. The proposed filter is based on parallel RLC circuit. The filter provides five output filter responses, namely, band-pass (BP, band-reject (BR, low-pass (LP, high-pass (HP, and all-pass (AP functions. The selection of each filter response can be done without the requirement of active and passive component matching condition. Furthermore, the natural frequency and quality factor are electronically controlled. Besides, the nonideal case is also investigated. The output voltage node exhibits low impedance. The experimental results can validate the theoretical analyses.

Decaffeination process characteristic of Robusta coffee in single column reactor using ethyl acetate solvent

Directory of Open Access Journals (Sweden)

Sukrisno Widyotomo

2009-08-01

Full Text Available Consumers drink coffee not as nutrition source, but as refreshment drink. For coffee consumers who have high tolerance for caffeine, coffee may warm up and refresh their bodies. High caffeine content in coffee beans may cause several complaints to consumers who are susceptible to caffeine. One of the efforts, for coffee market expansion is product diversification to decaffeinated coffee. Decaffeination process is one of process to reduce caffeine content from agricultural products. Indonesian Coffee and Cocoa Research Institute in collaboration with Bogor Agricultural University has developed a single column reactor for coffee beans decaffeination. The aim of this research is to study process characteristic of coffee decaffeination in single column reactor using ethyl acetate (C4H8O2 solvent. Treatments applicated in the research were time and temperature process. Temperature treatment were 50—60OC, 60—70OC, 70—80OC, 80—90OC and 90—100OC. Time treatment were 2 h, 4 h, 6 h, 8 h, 10 h, and 12 h Size of Robusta coffee beans used were less than 5.5 mm (A4, between 5.5 mm and 6.5 mm (A3, between 6.5 mm and 7.5 mm (A2, and more than 7.5 mm (A1. The result showed that decaffeination process with ethyl acetate solvent will be faster when its temperature was higher and smaller bean size. For bean size less than 5,5 mm, decaffeination process by 10% ethyl acetat can be done 8—10 hours in 90—100OC solvent temperature or 12 hours in 60—70OC solvent temperature for 0.3% caffein content. Organoleptic test showed that 90—100OC temperature solvent treatment decreased coffee flavor, which aroma, bitterness and body values were 1.9 each . Key words : Coffee, caffeine, decaffeination, quality, single column.

Acoustic characterization of a CANDU primary heat transport pump at the blade-passing frequency

International Nuclear Information System (INIS)

Rzentkowski, G.; Zbroja, S.

2000-01-01

In this paper, we examine the acoustics of a single-stage, double-volute CANDU heat transport pump based on a full-scale experimental investigation. We estimate the strength of source variables (acoustic pressure and velocity) and establish the pump characteristics as an acoustic source at the blade-passing frequency. We conduct this analysis by first assessing the resonance effects in the test loop, and then decomposing the measured signal into the components associated with pump action and loop acoustics with the use of a simple pump model. The pump model is based on a linear superposition of pressure wave transmission and excitation. The results of this analysis indicate that the pump source variables are nearly free of acoustic resonance effects in the test loop. The source pressure and velocity are each estimated at approximately 10 kPa (zero-to-peak). The results also indicate that the pump may act as both a pressure and a velocity source. At the loop resonance, the pump acoustic behavior is exclusively governed by the pressure term. This observation leads to the conclusion that the maximum amplification of pressure pulsations in a reactor heat transport system may be predicted by modeling the pump as a pressure source. (orig.)

Real-time digital signal recovery for a multi-pole low-pass transfer function system.

Science.gov (United States)

Lee, Jhinhwan

2017-08-01

In order to solve the problems of waveform distortion and signal delay by many physical and electrical systems with multi-pole linear low-pass transfer characteristics, a simple digital-signal-processing (DSP)-based method of real-time recovery of the original source waveform from the distorted output waveform is proposed. A mathematical analysis on the convolution kernel representation of the single-pole low-pass transfer function shows that the original source waveform can be accurately recovered in real time using a particular moving average algorithm applied on the input stream of the distorted waveform, which can also significantly reduce the overall delay time constant. This method is generalized for multi-pole low-pass systems and has noise characteristics of the inverse of the low-pass filter characteristics. This method can be applied to most sensors and amplifiers operating close to their frequency response limits to improve the overall performance of data acquisition systems and digital feedback control systems.

Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

Energy Technology Data Exchange (ETDEWEB)

Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

International Nuclear Information System (INIS)

Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu

2016-01-01

To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day -1 of air, 0.004%·day -1 of noble gas and 3.7×10 -5 %·day -1 of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m 3 ·hr -1 , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr -1 under the condition of 20 m·sec -1 wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor

Risk-informed design guidance for future reactor systems

International Nuclear Information System (INIS)

Delaney, Michael J.; Apostolakis, George E.; Driscoll, Michael J.

2005-01-01

Future reactor designs face an uncertain regulatory environment. It is anticipated that there will be some level of probabilistic insights in the regulations and supporting regulatory documents for Generation-IV nuclear reactors. Central to current regulations are design basis accidents (DBAs) and the general design criteria (GDC), which were established before probabilistic risk assessments (PRAs) were developed. These regulations implement a structuralist approach to safety through traditional defense in depth and large safety margins. In a rationalist approach to safety, accident frequencies are quantified and protective measures are introduced to make these frequencies acceptably low. Both approaches have advantages and disadvantages and future reactor design and licensing processes will have to implement a hybrid approach. This paper presents an iterative four-step risk-informed methodology to guide the design of future-reactor systems using a gas-cooled fast reactor emergency core cooling system as an example. This methodology helps designers to analyze alternative designs under potential risk-informed regulations and to anticipate design justifications the regulator may require during the licensing process. The analysis demonstrated the importance of common-cause failures and the need for guidance on how to change the quantitative impact of these potential failures on the frequency of accident sequences as the design changes. Deliberation is an important part of the four-step methodology because it supplements the quantitative results by allowing the inclusion in the design choice of elements such as best design practices and ease of online maintenance, which usually cannot be quantified. The case study showed that, in some instances, the structuralist and the rationalist approaches were inconsistent. In particular, GDC 35 treats the double-ended break of the largest pipe in the reactor coolant system with concurrent loss of offsite power and a single

Control rod for the operation of nuclear reactor

International Nuclear Information System (INIS)

Ishida, Hiromi

1987-01-01

Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

Degradation of benzodiazepines using water falling film dielectric barrier discharge reactor

Directory of Open Access Journals (Sweden)

Radulović Vesna M.

2017-01-01

Full Text Available Classical methods of wastewater treatment are often not suitable for the treatment of pharmaceutical waste. The previous studies have shown that the use of the advanced oxidation procedures (AOP can lead to a more efficient degradation of various biologically active compounds, which are active pharmaceutical ingredients of applied drugs. The aim of this paper is the application of the plasma technology on the degradation of a two active pharmaceutical ingredients (APIs, diazepam and alprazolam and the finished products (Bensedin® and Ksalol® using the dielectric barrier discharge (DBD reactor for AOP. We studied the degradation rate of these pharmaceuticals, depending on the number of passes through the reactor. This degradation method was efficient 61 % for diazepam and 95 % alprazolam. We also examined the influence of the pH adjustment between the passes of APIs through the DBD reactor. The degradation rate of APIs and the finished products was monitored by the high performance liquid chromatography (HPLC technique, using a photodiode array detector. The concentration of the dissolved ozone was determined using the iodometric procedure. [Project of the Serbian Ministry of Education, Science and Technological Development, Grant no. 172030

Performance of a palladium membrane reactor using a Ni catalyst for fusion fuel impurities processing

International Nuclear Information System (INIS)

Willms, R.S.; Wilhelm, R.; Okuno, K.

1994-01-01

The palladium membrane reactor (PNM) provides a means to recover hydrogen isotopes from impurities expected to be present in fusion reactor exhaust. This recovery is based on reactions such as water-gas shift and steam reforming for which conversion is equilibrium limited. By including a selectively permeable membrane such as Pd/Ag in the catalyst bed, hydrogen isotopes can be removed from the reacting environment, thus promoting the reaction to complete conversion. Such a device has been built and operated at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). For the reactions listed above, earlier study with this unit has shown that hydrogen single-pass recoveries approaching 100% can be achieved. It was also determined that a nickel catalyst is a feasible choice for use with a PMR appropriate for fusion fuel impurities processing. The purpose of this study was to systematically assess the performance of the PMR using a nickel catalyst over a range of temperatures, feed compositions and flowrates. Reactions which were studied are the water-gas shift reaction and steam reforming

Single-pass continuous-flow leach test of PNL 76-68 glass: some selected Bead Leach I results

International Nuclear Information System (INIS)

Coles, D.G.

1981-01-01

A single-pass continuous-flow leach test of PNL 76-68 glass beads (7 mm dia) was concluded after 420 days of uninterrupted operation. Variables included in the experimental matrix were flow-rate, leachant composition, and temperature. Analysis was conducted on all leachate samples for 237 Np and 239 Pu as well as a number of nonradioactive elements. Results indicated that flow-rate and leachant systematically affected the leach rate, but only slightly. Temperature effects were significant. Plutonium leach rate was lower at higher temperature suggesting that Pu sorption onto the beads was enhanced at the higher temperature. The range of leach rates for all analyzed elements (except Pu), at both temperature, at all three flow rates, and with all three leachant compositions varied only three orders of magnitude. The range of variables used in this experiment covered those expected in many proposed repository environments. The preliminary interpretation of the results also indicated that matrix dissolution may be the dominant leaching mechanism, at least for Np in bicarbonate leachant. Regardless of the leaching mechanism the importance of this study is that it bounds the effects of repository environments when the ground water is oxidizing and when it doesn't reach the waste form until the waste has cooled to ambient rock temperature

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

Graphite moderated reactor for thermoelectric generation

International Nuclear Information System (INIS)

Akazawa, Issei; Yamada, Akira; Mizogami, Yorikata

1998-01-01

Fuel rods filled with cladded fuel particles distributed and filled are buried each at a predetermined distance in graphite blocks situated in a reactor core. Perforation channels for helium gas as coolants are formed to the periphery thereof passing through vertically. An alkali metal thermoelectric power generation module is disposed to the upper lid of a reactor container while being supported by a securing receptacle. Helium gas in the coolant channels in the graphite blocks in the reactor core absorbs nuclear reaction heat, to be heated to a high temperature, rises upwardly by the reduction of the specific gravity, and then flows into an upper space above the laminated graphite block layer. Then the gas collides against a ceiling and turns, and flows down in a circular gap around the circumference of the alkali metal thermoelectric generation module. In this case, it transfers heat to the alkali metal thermoelectric generation module. (I.N.)

FY2016 ILAW Glass Corrosion Testing with the Single-Pass Flow-Through Method

Energy Technology Data Exchange (ETDEWEB)

Neeway, James J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Asmussen, Robert M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Parruzot, Benjamin PG [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cordova, Elsa [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Williams, Benjamin D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Leavy, Ian I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Stephenson, John R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); McElroy, Erin M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-04-21

The inventory of immobilized low-activity waste (ILAW) produced at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) will be disposed of at the near-surface, on-site Integrated Disposal Facility (IDF). When groundwater comes into contact with the waste form, the glass will corrode and radionuclides will be released into the near-field environment. Because the release of the radionuclides is dependent on the dissolution rate of the glass, it is important that the performance assessment (PA) model accounts for the dissolution rate of the glass as a function of various chemical conditions. To accomplish this, an IDF PA model based on Transition State Theory (TST) can be employed. The model is able to account for changes in temperature, exposed surface area, and pH of the contacting solution as well as the effect of silicon concentrations in solution, specifically the activity of orthosilicic acid (H4SiO4), whose concentration is directly linked to the glass dissolution rate. In addition, the IDF PA model accounts for the alkali-ion exchange process as sodium is leached from the glass and into solution. The effect of temperature, pH, H4SiO4 activity, and the rate of ion-exchange can be parameterized and implemented directly into the PA rate law model. The rate law parameters are derived from laboratory tests with the single-pass flow-through (SPFT) method. To date, rate law parameters have been determined for seven ILAW glass compositions, thus additional rate law parameters on a wider range of compositions will supplement the existing body of data for PA maintenance activities. The data provided in this report can be used by ILAW glass scientists to further the understanding of ILAW glass behavior, by IDF PA modelers to use the rate law parameters in PA modeling efforts, and by Department of Energy (DOE) contractors and decision makers as they assess the IDF PA program.

Humidity control device in a reactor container

International Nuclear Information System (INIS)

Aizawa, Motohiro; Igarashi, Hiroo; Osumi, Katsumi; Kimura, Takashi.

1986-01-01

Purpose: To provide a device capable of maintaining the inside of a container under high humidity circumstantial conditions causing less atmospheric corrosions, in order to prevent injuries due to atmospheric corrosions to smaller diameter stainless steel pipeways in the reactor container. Constitution: Stress corrosion cracks (SCC) to the smaller diameter stainless steel pipeways are caused dependent on the relative humidity and it is effective as the countermeasure against SCC to maintain the relative humidity at a low level less than 30 % or high level greater than 60 %. Based on the above findings, a humidity control device is disposed so as to maintain the relative humidity for the atmosphere within a reactor core on a higher humidity region. The device is adapted such that recycling gas in the dry-well coolant circuit is passed through an orifice to atomize the water introduced from feedwater pipe and introduce into a reactor core or such that the recycling gases in the dry-well cooling circuit are bubbled into water to remove chlorine gas in the reactor container gas thereby increasing the humidity in the reactor container. (Kamimura, M.)

A membrane-free, continuously feeding, single chamber up-flow biocatalyzed electrolysis reactor for nitrobenzene reduction

International Nuclear Information System (INIS)

Wang, Ai-Jie; Cui, Dan; Cheng, Hao-Yi; Guo, Yu-Qi; Kong, Fan-Ying; Ren, Nan-Qi; Wu, Wei-Min

2012-01-01

Highlights: ► A novel membrane-free up-flow biocatalyzed electrolysis reactor (UBER) was developed. ► Nitrobenzene as the mode of nitroaromatics was efficiently converted to aniline. ► The impact of phosphate buffer and acetate concentrations and power supplied were investigated. ► The prospects of UBER for the recalcitrant compound removal were discussed. - Abstract: A new bioelectrochemical system (BES), a membrane-free, continuous feeding up-flow biocatalyzed electrolysis reactor (UBER) was developed to reduce oxidative toxic chemicals to less- or non-toxic reduced form in cathode zone with oxidation of electron donor in anode zone. Influent was fed from the bottom of UBER and passed through cathode zone and then anode zone. External power source (0.5 V) was provided between anode and cathode to enhance electrochemical reactions. Granular graphite and carbon brush were used as cathode and anode, respectively. This system was tested for the reduction of nitrobenzene (NB) using acetate as electron donor and carbon source. The influent contained NB (50–200 mg L −1 ) and acetate (1000 mg L −1 ). NB was removed by up to 98% mainly in cathode zone. The anode potential maintained under −480 mV. The maximum NB removal rate was up to 3.5 mol m −3 TV d −1 (TV = total empty volume) and the maximum aniline (AN) formation rate was 3.06 mol m −3 TV d −1 . Additional energy required was less than 0.075 kWh mol −1 NB. The molar ratio of NB removed vs acetate consumed varied from 4.3 ± 0.4 to 2.3 ± 0.1 mol mol −1 . Higher influent phosphate or acetate concentration helped NB removal rate. NB could be efficiently reduced to AN as the power supplied of 0.3 V.

Experimental data and numerical predictions of a single-phase flow in a batch square stirred tank reactor with a rotating cylinder agitator

Science.gov (United States)

Escamilla-Ruíz, I. A.; Sierra-Espinosa, F. Z.; García, J. C.; Valera-Medina, A.; Carrillo, F.

2017-09-01

Single-phase flows in stirred tank reactors have useful characteristics for a wide number of industrial applications. Usually, reactors are cylindrical vessels and complex impeller designs, which are often highly energy consuming and produce complicated flow patterns. Therefore, a novel configuration consisting of a square stirred tank reactor is proposed in this study with potential advantages over conventional reactors. In the present work hydrodynamics and turbulence have been studied for a single-phase flow in steady state operating in batch condition. The flow was induced by drag from a rotating cylinder with two diameters. The effects of drag from the stirrer as well as geometrical parameters of the system on the hydrodynamic behavior were investigated using Computational Fluids Dynamics (CFD) and non-intrusive Laser Doppler Anemometry, (LDA). Data obtained from LDA measurements were used for the validation of the CFD simulations, and to detecting the macro-instabilities inside the tank, based on the time series analysis for three rotational speeds N = 180, 1000 and 2000 rpm. The numerical results revealed the formation of flow patterns and macro-vortex structures in the upper part of the tank as consequence of the Reynolds number and the stream discharge emanated from the cylindrical stirrer. Moreover, increasing the cylinder diameter has an impact on the number of recirculation loops as well as the energy consumption of the entire system showing better performance in the presence of turbulent flows.

Microfluidic Manufacturing of Polymeric Nanoparticles: Comparing Flow Control of Multiscale Structure in Single-Phase Staggered Herringbone and Two-Phase Reactors.

Science.gov (United States)

Xu, Zheqi; Lu, Changhai; Riordon, Jason; Sinton, David; Moffitt, Matthew G

2016-12-06

We compare the microfluidic manufacturing of polycaprolactone-block-poly(ethylene oxide) (PCL-b-PEO) nanoparticles (NPs) in a single-phase staggered herringbone (SHB) mixer and in a two-phase gas-liquid segmented mixer. NPs generated from two different copolymer compositions in both reactors and at three different flow rates, along with NPs generated using a conventional bulk method, are compared with respect to morphologies, dimensions, and internal crystallinities. Our work, the first direct comparison between alternate microfluidic NP synthesis methods, shows three key findings: (i) NP morphologies and dimensions produced in the bulk are different from those produced in a microfluidic mixer, whereas NP crystallinities produced in the bulk and in the SHB mixer are similar; (ii) NP morphologies, dimensions, and crystallinities produced in the single-phase SHB and two-phase mixers at the lowest flow rate are similar; and (iii) NP morphologies, dimensions, and crystallinities change with flow rate in the two-phase mixer but not in the single-phase SHB mixer. These findings provide new insights into the relative roles of mixing and shear in the formation and flow-directed processing of polymeric NPs in microfluidics, informing future reactor designs for manufacturing NPs of low polydispersity and controlled multiscale structure and function.

Present state of the liner of the reactor; Estado actual del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F; Raya A, R; Mazon R, R [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-07-15

When being presented to work the operation personnel of the reactor, on Monday January 10, 1983, they noticed that the reactor pool was overflowing of water and the floor of the room was partially flooded. The personnel proceeded to revise the feedwater systems to the pool, the Emergency Cooling System of the core and that of Water of Reinstatement, was found that the passing valve of this last it was lightly open. It was discovered that the water that was flooded in the floor of the room it came from the relief valves of the ports TW-1 and RW-2 and of three glides that were in the Thermal Column area. It was proceeded to lower the one level of water of the pool to their normal position and it was clean the water flooded in the salts. (Author)

Nuclear reactor spacer assembly

International Nuclear Information System (INIS)

Anthony, A.J.; Groves, M.D.

1979-01-01

A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

International Nuclear Information System (INIS)

Wittenbrock, N.G.

1982-01-01

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

Development of SiC Nanoparticles and Second Phases Synergistically Reinforced Mg-Based Composites Processed by Multi-Pass Forging with Varying Temperatures

Directory of Open Access Journals (Sweden)

Kaibo Nie

2018-01-01

Full Text Available In this study, SiC nanoparticles were added into matrix alloy through a combination of semisolid stirring and ultrasonic vibration while dynamic precipitation of second phases was obtained through multi-pass forging with varying temperatures. During single-pass forging of the present composite, as the deformation temperature increased, the extent of recrystallization increased, and grains were refined due to the inhibition effect of the increasing amount of dispersed SiC nanoparticles. A small amount of twins within the SiC nanoparticle dense zone could be found while the precipitated phases of Mg17Al12 in long strips and deformation bands with high density dislocations were formed in the particle sparse zone after single-pass forging at 350 °C. This indicated that the particle sparse zone was mainly deformed by dislocation slip while the nanoparticle dense zone may have been deformed by twinning. The yield strength and ultimate tensile strength of the composites were gradually enhanced through increasing the single-pass forging temperature from 300 °C to 400 °C, which demonstrated that initial high forging temperature contributed to the improvement of the mechanical properties. During multi-pass forging with varying temperatures, the grain size of the composite was gradually decreased while the grain size distribution tended to be uniform with reducing the deformation temperature and extending the forging passes. In addition, the amount of precipitated second phases was significantly increased compared with that after multi-pass forging under a constant temperature. The improvement in the yield strength of the developed composite was related to grain refinement strengthening and Orowan strengthening resulting from synergistical effect of the externally applied SiC nanoparticles and internally precipitated second phases.

Improvements in or relating to gripping means for handling nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Batjukov, V.I.; Vjugov, O.N.; Fadeev, A.I.; Shkhian, T.G.

1980-01-01

A gripping means for handling fuel assemblies, the heads of which are internally recessed to receive gripping jaws, forms part of a reactor refuelling machine and is telescopically accommodated within a manipulator tube of the machine. A through hole is provided to allow cooling medium to be passed through the fuel assemblies to remove afterheat when the gripping means is used to transfer assemblies from a reactor core to spent fuel storage sockets. (author)

Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

International Nuclear Information System (INIS)

Powers, J.J.; Worrall, A.; Gehin, J.C.; Harrison, T.J.; Sunny, E.E.

2013-01-01

Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

Analyses of Decrease in Reactor Coolant Flow Rate in SMART

International Nuclear Information System (INIS)

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

2011-01-01

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

Collector Efficiency in Downward-Type Double-Pass Solar Air Heaters with Attached Fins and Operated by External Recycle

Directory of Open Access Journals (Sweden)

Chii-Dong Ho

2012-07-01

Full Text Available The collector efficiency in a downward-type double-pass external-recycle solar air heater with fins attached on the absorbing plate has been investigated theoretically. Considerable improvement in collector efficiency is obtainable if the collector is equipped with fins and the operation is carried out with an external recycle. Due to the recycling, the desirable effect of increasing the heat transfer coefficient compensates for the undesirable effect of decreasing the driving force (temperature difference of heat transfer, while the attached fins provide an enlarged heat transfer area. The order of performances in the devices of same size is: double pass with recycle and fins > double pass with recycle but without fins > single pass without recycle and fins.

COD fractions of leachate from aerobic and anaerobic pilot scale landfill reactors

International Nuclear Information System (INIS)

Bilgili, M. Sinan; Demir, Ahmet; Akkaya, Ebru; Ozkaya, Bestamin

2008-01-01

One of the most important problems with designing and maintaining a landfill is managing leachate that generated when water passes through the waste. In this study, leachate samples taken from aerobic and anaerobic landfill reactors operated with and without leachate recirculation are investigated in terms of biodegradable and non-biodegradable fractions of COD. The operation time is 600 days for anaerobic reactors and 250 days for aerobic reactors. Results of this study show that while the values of soluble inert COD to total COD in the leachate of aerobic landfill with leachate recirculation and aerobic dry reactors are determined around 40%, this rate was found around 30% in the leachate of anaerobic landfill with leachate recirculation and traditional landfill reactors. The reason for this difference is that the aerobic reactors generated much more microbial products. Because of this condition, it can be concluded that total inert COD/total COD ratios of the aerobic reactors were 60%, whereas those of anaerobic reactors were 50%. This study is important for modeling, design, and operation of landfill leachate treatment systems and determination of discharge limits

Device Performance Improvement of Double-Pass Wire Mesh Packed Solar Air Heaters under Recycling Operation Conditions

Directory of Open Access Journals (Sweden)

Chii-Dong Ho

2016-01-01

Full Text Available The improvement of device performance of a recycling solar air heater featuring a wire mesh packing was investigated experimentally and theoretically. The application of the wire mesh packing and recycle-effect concept to the present study were proposed aiming to strengthen the convective heat-transfer coefficient due to increased turbulence. Comparisons were made among different designs, including the single-pass, flat-plate double-pass and recycling double-pass wire mesh packed operations. The collector efficiency of the recycling double-pass wire mesh packed solar air heater was much higher than that of the other configurations for various recycle ratios and mass flow rates scenarios. The power consumption increment due to implementing wire mesh in solar air heaters was also discussed considering the economic feasibility. A fairly good agreement between theoretical predictions and experimental measurements was achieved with an analyzed error of 1.07%–9.32%.

Low-pass sequencing for microbial comparative genomics

Directory of Open Access Journals (Sweden)

Kennedy Sean

2004-01-01

Full Text Available Abstract Background We studied four extremely halophilic archaea by low-pass shotgun sequencing: (1 the metabolically versatile Haloarcula marismortui; (2 the non-pigmented Natrialba asiatica; (3 the psychrophile Halorubrum lacusprofundi and (4 the Dead Sea isolate Halobaculum gomorrense. Approximately one thousand single pass genomic sequences per genome were obtained. The data were analyzed by comparative genomic analyses using the completed Halobacterium sp. NRC-1 genome as a reference. Low-pass shotgun sequencing is a simple, inexpensive, and rapid approach that can readily be performed on any cultured microbe. Results As expected, the four archaeal halophiles analyzed exhibit both bacterial and eukaryotic characteristics as well as uniquely archaeal traits. All five halophiles exhibit greater than sixty percent GC content and low isoelectric points (pI for their predicted proteins. Multiple insertion sequence (IS elements, often involved in genome rearrangements, were identified in H. lacusprofundi and H. marismortui. The core biological functions that govern cellular and genetic mechanisms of H. sp. NRC-1 appear to be conserved in these four other halophiles. Multiple TATA box binding protein (TBP and transcription factor IIB (TFB homologs were identified from most of the four shotgunned halophiles. The reconstructed molecular tree of all five halophiles shows a large divergence between these species, but with the closest relationship being between H. sp. NRC-1 and H. lacusprofundi. Conclusion Despite the diverse habitats of these species, all five halophiles share (1 high GC content and (2 low protein isoelectric points, which are characteristics associated with environmental exposure to UV radiation and hypersalinity, respectively. Identification of multiple IS elements in the genome of H. lacusprofundi and H. marismortui suggest that genome structure and dynamic genome reorganization might be similar to that previously observed in the

Fabrication of seamless calandria tubes by cold pilgering route using 3-pass and 2-pass schedules

Science.gov (United States)

Saibaba, N.

2008-12-01

Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90-95. Such tubes are conventionally produced by the 'welded route', which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the 'seamless route' which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties.

Fabrication of seamless calandria tubes by cold pilgering route using 3-pass and 2-pass schedules

International Nuclear Information System (INIS)

Saibaba, N.

2008-01-01

Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90-95. Such tubes are conventionally produced by the 'welded route', which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the 'seamless route' which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties

Coupling membrane pervaporation with a fixed-bed reactor for enhanced esterification of oleic acid with ethanol

International Nuclear Information System (INIS)

Han, Ying; Lv, Enmin; Ma, Lingling; Lu, Jie; Chen, Kexun; Ding, Jincheng

2015-01-01

Highlights: • The reactor coupling membrane pervaporation with a fixed-bed reactor was studied. • The factors effecting the esterification of oleic acid were investigated. • NaA zeolite membrane was used for dehydration in the coupled reactor. - Abstract: Process intensification through membrane pervaporation (PV) integrated with a fixed-bed reactor could be successfully applied to the esterification of oleic acid and ethanol, which is a crucial step in the biodiesel synthesis using waste oil and grease as resource. The properties of the NaA zeolite membrane such as structure, formulation and separation were investigated by scanning electronic microscopy–energy dispersive spectrometry (SEM–EDS), X-ray diffractometry (XRD) and PV dehydration. Results showed that the NaA zeolite membrane had good separating property for removing water from the organics mixture. The operating conditions were optimized as the ethanol to oleic acid molar ratio of 15:1, feedstock flow rate of 1.0 ml/min, reaction temperature of 80.0 °C and catalyst bed height of 132 mm. The final conversion of oleic acid increased from 84.23% to 87.18% by PV using the NaA zeolite membrane at 24.0 h of operation. The membrane showed good PV performance after used for eight successive runs in the PV-assisted esterification. The resin exhibited a much high catalytic activity and operation stability after used for 100 h in the consecutive single pass fixed-bed esterification.

Tunable First-Order Resistorless All-Pass Filter with Low Output Impedance

Directory of Open Access Journals (Sweden)

Parveen Beg

2014-01-01

Full Text Available This paper presents a voltage mode cascadable single active element tunable first-order all-pass filter with a single passive component. The active element used to realise the filter is a new building block termed as differential difference dual-X current conveyor with a buffered output (DD-DXCCII. The filter is thus realized with the help of a DD-DXCCII, a capacitor, and a MOS transistor. By exploiting the low output impedance, a higher order filter is also realized. Nonideal and parasitic study is also carried out on the realised filters. The proposed DD-DXCCII filters are simulated using TSMC the 0.25 µm technology.

Tunable first-order resistorless all-pass filter with low output impedance.

Science.gov (United States)

Beg, Parveen

2014-01-01

This paper presents a voltage mode cascadable single active element tunable first-order all-pass filter with a single passive component. The active element used to realise the filter is a new building block termed as differential difference dual-X current conveyor with a buffered output (DD-DXCCII). The filter is thus realized with the help of a DD-DXCCII, a capacitor, and a MOS transistor. By exploiting the low output impedance, a higher order filter is also realized. Nonideal and parasitic study is also carried out on the realised filters. The proposed DD-DXCCII filters are simulated using TSMC the 0.25 µm technology.

Oscillatory flow chemical reactors

Directory of Open Access Journals (Sweden)

Slavnić Danijela S.

2014-01-01

Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn

International Nuclear Information System (INIS)

Talamo, Alberto; Ji, Wei; Cetnar, Jerzy; Gudowski, Waclaw

2006-01-01

Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides

Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)]. E-mail: alby@anl.gov; Ji, Wei [University of Michigan, Bonisteel Boulevard 2355, Ann Arbor, MI 48109-2104 (United States); Cetnar, Jerzy [AGH-University of Science and Technology, Al. Mickiewicza 30 Cracow (Poland); Gudowski, Waclaw [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)

2006-09-15

Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Determination of the Clean Air Delivery Rate (CADR of Photocatalytic Oxidation (PCO Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part I: Theoretical Considerations

Directory of Open Access Journals (Sweden)

Éric Dumont

2017-03-01

Full Text Available This study demonstrated that a laboratory-scale recirculation closed-loop reactor can be an efficient technique for the determination of the Clean Air Delivery Rate (CADR of PhotoCatalytic Oxidation (PCO air purification devices. The recirculation closed-loop reactor was modeled by associating equations related to two ideal reactors: one is a perfectly mixed reservoir and the other is a plug flow system corresponding to the PCO device itself. Based on the assumption that the ratio between the residence time in the PCO device and the residence time in the reservoir τP/τR tends to 0, the model highlights that a lab closed-loop reactor can be a suitable technique for the determination of the efficiency of PCO devices. Moreover, if the single-pass removal efficiency is lower than 5% of the treated flow rate, the decrease in the pollutant concentration over time can be characterized by a first-order decay model in which the time constant is proportional to the CADR. The limits of the model are examined and reported in terms of operating conditions (experiment duration, ratio of residence times, and flow rate ranges.

Possibilities of the fish pass restoration

Science.gov (United States)

Čubanová, Lea

2018-03-01

According to the new elaborated methodology of the Ministry of Environment of the Slovak Republic: Identification of the appropriate fish pass types according to water body typology (2015) each barrier on the river must be passable. On the barriers or structures without fish passes new ones should be design and built and on some water structures with existed but nonfunctional fish passes must be realized reconstruction or restoration of such objects. Assessment should be done in terms of the existing migratory fish fauna and hydraulic conditions. Fish fauna requirements resulting from the ichthyological research of the river section with barrier. Hydraulic conditions must than fulfil these requirements inside the fish pass body.

Enhanced Hydrogen Production Integrated with CO2 Separation in a Single-Stage Reactor

Energy Technology Data Exchange (ETDEWEB)

Mahesh Iyer; Himanshu Gupta; Danny Wong; Liang-Shih Fan

2005-09-30

Hydrogen production from coal gasification can be enhanced by driving the equilibrium limited Water Gas Shift reaction forward by incessantly removing the CO{sub 2} by-product via the carbonation of calcium oxide. This project aims at using the OSU patented high-reactivity mesoporous precipitated calcium carbonate sorbent for removing the CO{sub 2} product. Preliminary experiments demonstrate the show the superior performance of the PCC sorbent over other naturally occurring calcium sorbents. Gas composition analyses show the formation of 100% pure hydrogen. Novel calcination techniques could lead to smaller reactor footprint and single-stage reactors that can achieve maximum theoretical H{sub 2} production for multicyclic applications. Sub-atmospheric calcination studies reveal the effect of vacuum level, diluent gas flow rate, thermal properties of the diluent gas and the sorbent loading on the calcination kinetics which play an important role on the sorbent morphology. Steam, which can be easily separated from CO{sub 2}, is envisioned to be a potential diluent gas due to its enhanced thermal properties. Steam calcination studies at 700-850 C reveal improved sorbent morphology over regular nitrogen calcination. A mixture of 80% steam and 20% CO{sub 2} at ambient pressure was used to calcine the spent sorbent at 820 C thus lowering the calcination temperature. Regeneration of calcium sulfide to calcium carbonate was achieved by carbonating the calcium sulfide slurry by bubbling CO{sub 2} gas at room temperature.

Concept and designs of new-generation fast reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.

1993-01-01

This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

Leak detector for a steam generator in FBR type reactors

International Nuclear Information System (INIS)

Miyaji, Nobuyoshi.

1979-01-01

Purpose: To facilitate maintenance for liquid leak detectors such as exchange of nickel membrane sensors during operation in a sodium-cooled fbr type reactor. Constitution: A pipeway capable of supplying a cover gas such as argon into the cylinder of a hydrogen detector containing a nickel membrane sensor is provided in a liquid leak detector constituting a part of a by-pass loop. The pipeway is also adapted to be evacuated. A pipeway and a small sodium tank for drain use are provided on the side of the by-pass loop near valves. Then, after closing the inlet and outlet valves to disconnect the by-pass loop from the sodium main pipeway, the cover gas is supplied to drive liquid sodium to the drain tank. After the drain of the liquid sodium, the sensor can be replaced. (Ikeda, J.)

WebPASS ICASS (HR Personnel Management)

Data.gov (United States)

US Agency for International Development — WebPASS Joint Administrative Support Platforms Post Administrative Software Suite - U.S. Department of State Executive Officers application suite. Web.PASS is the...

Experience of on-site disposal of production uranium-graphite nuclear reactor.

Science.gov (United States)

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

Cleaning device for recycling pump motor cooling system in nuclear reactor

International Nuclear Information System (INIS)

Katayama, Kenjiro; Kondo, Takahisa; Shindo, Kenjiro; Akimoto, Jun.

1996-01-01

The cleaning device of the present invention comprises a cleaning water supply pump, a filter for filtering the cleaning water and a cap member for isolating the inside of a motor casing from the inside of a reactor pressure vessel. A motor in the motor casing and a pump in the reactor pressure vessel are removed, the cap member is attached to the upper end of the motor casing to isolate the inside of the motor casing from the inside of the reactor pressure vessel. If the cleaning water supply pump is operated in this state, the cleaning water flows from a returning pipeline for cooling water circulation, connected to the motor casing to supply pipelines through a heat exchange and is discharged. The discharged water passes through a filter and is sent again, as the cleaning water, to the cleaning water supply pump. With such procedures, the recycling pump motor cooling system in the BWR type reactor can be cleaned without disposing a cyclone separator and irrespective of presence or absence of reactor coolants in the reactor pressure vessel. (I.N.)

Message passing for quantified Boolean formulas

International Nuclear Information System (INIS)

Zhang, Pan; Ramezanpour, Abolfazl; Zecchina, Riccardo; Zdeborová, Lenka

2012-01-01

We introduce two types of message passing algorithms for quantified Boolean formulas (QBF). The first type is a message passing based heuristics that can prove unsatisfiability of the QBF by assigning the universal variables in such a way that the remaining formula is unsatisfiable. In the second type, we use message passing to guide branching heuristics of a Davis–Putnam–Logemann–Loveland (DPLL) complete solver. Numerical experiments show that on random QBFs our branching heuristics give robust exponential efficiency gain with respect to state-of-the-art solvers. We also manage to solve some previously unsolved benchmarks from the QBFLIB library. Apart from this, our study sheds light on using message passing in small systems and as subroutines in complete solvers

Oil price pass-through into inflation

International Nuclear Information System (INIS)

Chen, Shiu-Sheng

2009-01-01

This paper uses data from 19 industrialized countries to investigate oil price pass-through into inflation across countries and over time. A time-varying pass-through coefficient is estimated and the determinants of the recent declining effects of oil shocks on inflation are investigated. The appreciation of the domestic currency, a more active monetary policy in response to inflation, and a higher degree of trade openness are found to explain the decline in oil price pass-through. (author)

A Novel Dual-Stage Hydrothermal Flow Reactor

DEFF Research Database (Denmark)

Hellstern, Henrik Christian; Becker, Jacob; Hald, Peter

2015-01-01

The dual-stage reactor is a novel continuous flow reactor with two reactors connected in series. It is designed for hydrothermal flow synthesis of nanocomposites, in which a single particle consists of multiple materials. The secondary material may protect the core nanoparticle from oxidation....... The dual-stage reactor combines the ability to produce advanced materials with an upscaled capacity in excess of 10 g/hour (dry mass). TiO2 was synthesized in the primary reactor and reproduced previous results. The dual-stage capability was succesfully demonstrated with a series of nanocomposites incl. Ti...

VIPRE modeling of VVER-1000 reactor core for DNB analyses

Energy Technology Data Exchange (ETDEWEB)

Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

1995-09-01

Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

Influences of crystallographic orientations on deformation mechanism and grain refinement of Al single crystals subjected to one-pass equal-channel angular pressing

International Nuclear Information System (INIS)

Han, W.Z.; Zhang, Z.F.; Wu, S.D.; Li, S.X.

2007-01-01

The influences of crystallographic orientations on the evolution of dislocation structures and the refinement process of sub-grains in Al single crystals processed by one-pass equal-channel angular pressing (ECAP) were systematically investigated by means of scanning electron microscopy, electron backscatter diffraction and transmission electron microscopy. Three single crystals with different orientations, denoted as crystal I, crystal II and crystal III, were specially designed according to the shape of the ECAP die. For crystal I, its insert direction is parallel to [1 1 0] and its extrusion direction is parallel to [1-bar11]. For crystal II, the (1-bar11) plane is located parallel to the intersection plane of the ECAP die, and the [1 1 0] direction is along the general shear direction on the intersection plane. For crystal III, the (1-bar11) plane is laid on the plane perpendicular to the intersection of the ECAP die, and the [1 1 0] direction is vertical to the general shear direction. For crystal I, abundant cell block structures with multi-slip characters were formed, and they should be induced by four symmetric slip systems, while for crystal II, there are two sets of sub-grain structures with higher misorientation, making an angle of ∼70 deg., which can be attributed to the interactions of the two asymmetric primary slip planes, whereas for crystal III, only one set of ribbon structures was parallel to the traces of (1-bar11) with the lowest misorientation angle among the three single crystals, which should result from the homogeneous slip on the primary slip plane. The different microstructural features of the three single crystals provide clear experimental evidence that the microstructures and misorientation evolution are strongly affected by the crystallographic orientation or by the interaction between shear deformation imposed by the ECAP die and the intrinsic slip deformation of the single crystals. Based on the experimental results and the

Heating control system for nuclear reactor

International Nuclear Information System (INIS)

Shinohara, Kaoru.

1981-01-01

Purpose: To automatically control reactor heating while keeping the condition of temperature rising rate by determining the deviations based on the reactor water temperature, the aimed temperature and the aimed temperature rising rate and operating control rods. Constitution: Actual temperature in the reactor is measured by a temperature detector and compared with a value from a setter to determine the temperature deviation. While on the other hand, the rising rate for the measured temperature is calculated in a differentiator and compared with a value from a setter to determine the deviation, which is passed through an integrator to calculate the deviation for the temperature rising rate. The signals for the temperature deviation and the temperature rising rate deviation are selected in a lower value preference circuit and the operation amount for the control rod is judged in a control rod operation judging section depending on the deviation amount. The control rod to be operated is determined in a sequence control section for the selection of control rod. The control rod selected and the direction of the operation are displayed on a display and the selected control rod is automatically driven by a control rod drives to thereby carry our reactor heating. (Furukawa, Y.)

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Directory of Open Access Journals (Sweden)

Ternovykh Mikhail

2017-01-01

Full Text Available Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Science.gov (United States)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Optimization of band-pass filtering parameters of a Raman lidar detecting atmospheric water vapor

International Nuclear Information System (INIS)

Cao, Kai-Fa; Hu, Shun-Xing; Wang, Ying-jian

2012-01-01

It is very important for daytime Raman lidar measurement of water vapor to determine the parameters of a band-pass filter, which are pertinent to the lidar signal to noise ratio (SNR). The simulated annealing (SA) algorithm method has an advantage in finding the extremum of a certain cost function. In this paper, the Raman spectrum of water vapor is simulated and then a first realization of a simulated annealing algorithm in the optimization of a band-pass filter of a Raman lidar system designed to detect daytime water vapor is presented. The simulated results indicate that the narrow band-pass filter has higher SNR than the wide filter does but there would be an increase in the temperature sensitivity of a narrowband Raman water vapor lidar in the upper troposphere. The numerical simulation indicates that the magnitude of the temperature dependent effect can reach 3.5% or more for narrow band-pass Raman water vapor measurements so it is necessary to consider a new water vapor Raman lidar equation that permits the temperature sensitivity of these equations to be confined to a single term. (paper)

N-isopropyl-[123I]p-iodoamphetamine: single-pass brain uptake and washout; binding to brain synaptosomes; and localization in dog and monkey brain

International Nuclear Information System (INIS)

Winchell, H.S.; Horst, W.D.; Braun, L.; Oldendorf, W.H.; Hattner, R.; Parker, H.

1980-01-01

The kinetics of N-isopropyl-p-[ 123 I]iodoamphetamine in rat brains were determined by serial measurements of brain uptake index (BUI) after intracarotid injection; also studied were its effects on amine uptake and release in rat's brain cortical synaptosomes; and its in vivo distribution in the dog and monkey. No specific localization in brain nuclei of the dog was seen, but there was progressive accumulation in the eyes. Rapid initial brain uptake in the ketamine-sedated monkey was noted, and further slow brain uptake occurred during the next 20 min but without retinal localization. High levels of brain activity were maintained for several hours. The quantitative initial single-pass clearance of the agent in the brain suggests its use in evaluation of regional brain perfusion. Its interaction with brain amine-binding sites suggests its possible application in studies of cerebral amine metabolism

Investigation reactor D-2201 polypropylene production unit using nuclear technique

International Nuclear Information System (INIS)

Wibisono; Sugiharto; Jefri Simanjuntak

2016-01-01

D-2201 reactor is a unit in the polypropylene production process at Pertamina Refinery Unit III Plaju. Reactor with a capacity of 45 kilo liter is not operated in normal operation condition. The validity of liquid level indicator on the unit is doubtful when refers to the production quality. Gamma source of 150 mCi Cobalt-60 and a scintillation detector had been used to scan the outer wall of the reactor to detect the liquid level during operation with a capacity of 40 %. Measurements were made along the reactor walls with 25 mm scan resolution and 5 seconds time sampling. Experiment result shows that the liquid level at the position of 40 % and at normal level position are not observed. Investigation did not find the liquid level above normal. D-2201 is diagnose not normal operating condition diagnosed with liquid abundant passed the recommended limits. Investigation advised to repair or to calibrate the liquid level indicator which is currently installed. (author)

Annual harvests of Corbicula populations prevent clogging of nuclear reactor heat exchangers

International Nuclear Information System (INIS)

Harvey, R.S.

1983-01-01

An annual program for removal of millions of Corbicula from upstream cooling water basins has prevented reclogging of nuclear reactor heat exchanger distributor plates at the Savannah River Plant during the past seven years. There are nine 32-megaliter basins in the three operating reactor areas where some settling of particulates occurs before cooling water is passed through screens in route to heat exchangers. Annual cleanings keep silt/clam substrate levels low and clam sizes small. Data are presented on the size/age distribution for clams recolonizing basins between cleanings

RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

2003-04-01

The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

Neighbourhood-consensus message passing and its potentials in image processing applications

Science.gov (United States)

Ružic, Tijana; Pižurica, Aleksandra; Philips, Wilfried

2011-03-01

In this paper, a novel algorithm for inference in Markov Random Fields (MRFs) is presented. Its goal is to find approximate maximum a posteriori estimates in a simple manner by combining neighbourhood influence of iterated conditional modes (ICM) and message passing of loopy belief propagation (LBP). We call the proposed method neighbourhood-consensus message passing because a single joint message is sent from the specified neighbourhood to the central node. The message, as a function of beliefs, represents the agreement of all nodes within the neighbourhood regarding the labels of the central node. This way we are able to overcome the disadvantages of reference algorithms, ICM and LBP. On one hand, more information is propagated in comparison with ICM, while on the other hand, the huge amount of pairwise interactions is avoided in comparison with LBP by working with neighbourhoods. The idea is related to the previously developed iterated conditional expectations algorithm. Here we revisit it and redefine it in a message passing framework in a more general form. The results on three different benchmarks demonstrate that the proposed technique can perform well both for binary and multi-label MRFs without any limitations on the model definition. Furthermore, it manifests improved performance over related techniques either in terms of quality and/or speed.

A water inner circulation device for a reactor vessel

International Nuclear Information System (INIS)

Eriksson, O.

1976-01-01

A water inner circulation device for a reactor vessel comprising a pump mounted in the reactor vessel and driven by a water-cooled electric motor mounted in a housing outside the reactor vessel, the shaft of the pump passing through the reactor-vessel bottom and being coupled to the motor shaft in a member mechanically connected to the bottom of the reactor vessel in the vicinity of the motor housing, the pump shaft being surrounded by a resilient sealing ring, the reactor vessel communicating with the cooling channels of the pump, when the latter is operating, via a slot surrounding the pump hollow cylindrical shaft, characterized in that the slot inner end is used for/forming a circular space surrounding the pump shaft and surrounded by the motorhousing, in which is coaxially mounted a separating cylindral wall, the upper edge of which is tightly applied against the inner wall of the motor-housing to which it is fastened vertically, the inner surface of said wall being turned towards the outer surface of a circular packing-box, the outer surface of said separating wall constituting a separating radical inner surface for a circular chamber through which flow the motor cooling water. (author)

Hanford spent fuel inventory baseline

International Nuclear Information System (INIS)

Bergsman, K.H.

1994-01-01

This document compiles technical data on irradiated fuel stored at the Hanford Site in support of the Hanford SNF Management Environmental Impact Statement. Fuel included is from the Defense Production Reactors (N Reactor and the single-pass reactors; B, C, D, DR, F, H, KE and KW), the Hanford Fast Flux Test Facility Reactor, the Shipping port Pressurized Water Reactor, and small amounts of miscellaneous fuel from several commercial, research, and experimental reactors

Improvements to the sodium supply system of a nuclear reactor core

International Nuclear Information System (INIS)

Chevallier, Rene; Marchais, Christian.

1981-01-01

This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

Decommissioning of fast reactors after sodium draining

International Nuclear Information System (INIS)

2009-11-01

Acknowledging the importance of passing on knowledge and experience, as well mentoring the next generation of scientists and engineers, and in response to expressed needs by Member States, the IAEA has undertaken concrete steps towards the implementation of a fast reactor data retrieval and knowledge preservation initiative. Decommissioning of fast reactors and other sodium bearing facilities is a domain in which considerable experience has been accumulated. Within the framework and drawing on the wide expertise of the Technical Working Group on Fast Reactors (TWG-FR), the IAEA has initiated activities aiming at preserving the feedback (lessons learned) from this experience and condensing those to technical recommendations on fast reactor design features that would ease their decommissioning. Following a recommendation by the TWG-FR, the IAEA had convened a topical Technical Meeting (TM) on 'Operational and Decommissioning Experience with Fast Reactors', hosted by CEA, Centre d'Etudes de Cadarache, France, from 11 to 15 March 2002 (IAEA-TECDOC- 1405). The participants in that TM exchanged detailed technical information on fast reactor operation and decommissioning experience with various sodium cooled fast reactors, and, in particular, reviewed the status of the various decommissioning programmes. The TM concluded that the decommissioning of fast reactors to reach safe enclosure presented no major difficulties, and that this had been accomplished mainly through judicious adaptation of processes and procedures implemented during the reactor operation phase, and the development of safe sodium waste treatment processes. However, the TM also concluded that, on the path to achieving total dismantling, challenges remain with regard to the decommissioning of components after sodium draining, and suggested that a follow-on TM be convened, that would provide a forum for in-depth scientific and technical exchange on this topic. This publication constitutes the Proceedings of

Raising distillate selectivity and catalyst life time in Fischer-Tropsch synthesis by using a novel dual-bed reactor

International Nuclear Information System (INIS)

Tavasoli, A.; Sadaghiani, K.; Khodadadi, A. A.; Mortazavi, Y.

2007-01-01

In a novel dual bed reactor Fischer-Tropsch synthesis was studied by using two diff rent cobalt catalysts. An alkali-promoted cobalt catalyst was used in the first bed of a fixed-bed reactor followed by a Raiment promoted cobalt catalyst in the second bed. The activity, product selectivity and accelerated deactivation of the system were assessed and compared with a conventional single bed reactor system. The methane selectivity in the dual-bed reactor was about 18.9% less compared to that of the single-bed reactor. The C 5+ selectivity for the dual-bed reactor was 10.9% higher than that of the single-bed reactor. Accelerated deactivation of the catalysts in the dual-bed reactor was 42% lower than that of the single-bed reactor. It was revealed that the amount of catalysts activity recovery after regeneration at 400 d eg C in the dual-bed system is higher than that of the single-bed system

Comment on “Voltage-Mode All-Pass Filters Including Minimum Component Count Circuits”

OpenAIRE

Lahiri, Abhirup

2009-01-01

This comment is related to the recently published article “Active and Passive Electronic Components” by S. Maheshwari (2007), which presents single current differencing buffered amplifier (CDBA) and current-controlled current differencing buffered amplifier- (CC-CDBA-) based first-order voltage-mode (VM) all-pass filtering (APF) sections. The paper is reviewed, and additional first-order APF realizations have been proposed.

Welding of components of primary circuits of nuclear reactors in FRG

International Nuclear Information System (INIS)

Pehtts, P.; Iversen, K.

1979-01-01

Welding materials and methods, surfacing and soldering, applied when assembling nuclear reactors in the Federal Republic of Germany, are considered. It is noted that reactor vessel flux two-pass surfacing is mainly carried out, using the band electrode. The austenitic steel serves as filler material. Vessels are welded using electroslag flux method and nonconsumable electrodes. Tube plates claddina and tube welding during steam generator production are made by flux surfacing and inert gas shielded using nonconsumable electrode. When assembling fuel elements high temperature soldering with the solders, containing no boron of the Ni-Cr-Si and Ni-Cr-P systems is used

Aeration Strategies To Mitigate Nitrous Oxide Emissions from Single-Stage Nitritation/Anammox Reactors

DEFF Research Database (Denmark)

Domingo Felez, Carlos; Mutlu, A. Gizem; Jensen, Marlene Mark

2014-01-01

Autotrophic nitrogen removal is regarded as a resource efficient process to manage nitrogen-rich residual streams. However, nitrous oxide emissions of these processes are poorly documented and strategies to mitigate emissions unknown. In this study, two sequencing batch reactors performing single...... was noted when the duration of aeration was increased while decreasing air flow rate (10.9 +/- 3.2% Delta N2O/Delta TN). The extant ammonium oxidation activity (mgNH(4)(+)-N/gVSS.min) positively correlated with the specific N2O production rate (mgN(2)O-N/gVSS.min) of the systems. Operating under conditions......-stage nitritation/anammox were operated under different aeration strategies, gradually adjusted over six months. At constant but limiting oxygen loading, synthetic reject water was fed (0.75g-N/L.d) and high nitrogen removal efficiencies (83 +/- 5 and 88 +/- 2%) obtained. Dynamics of liquid phase nitrous (N2O...

Analytical and Experimental Study of Recycling Baffled Double-Pass Solar Air Heaters with Attached Fins

Directory of Open Access Journals (Sweden)

Chun Sheng Lin

2013-03-01

Full Text Available The study of the heat transfer of solar air heaters with a new design using an absorbing plate with fins and baffles, which facilitate the recycling of flowing air, is reported. The mathematical formulation and analytical analysis for such a recyclic baffled double-pass solar air heater were developed theoretically. The performance of the device was studied experimentally as well. The theoretical predicted and experimental results were compared with another design, i.e., a downward-type single-pass solar air heater without recycle and double-pass operations reported in our previous work. Significant improvement in heat-transfer efficiency is achieved with the baffle and fin design due to the recycling heating and the extended heat transfer area. The effects of mass flow rate and recycle ratio on the heat-transfer efficiency enhancement as well as on the power consumption increment are also discussed.

Data acquisition and processing system for reactor noise analysis

International Nuclear Information System (INIS)

Costa Oliveira, J.; Morais Da Veiga, C.; Forjaz Trigueiros, D.; Pombo Duarte, J.

1975-01-01

A data acquisition and processing system for reactor noise analysis by time correlation methods is described, consisting in one to four data feeding channels (transducer, associated electronics and V/f converter), a sampling unit, a landline transmission system and a PDP 15 computer. This system is being applied to study the kinetic parameters of the 'Reactor Portugues de Investigacao', a swimming-pool 1MW reactor. The main features that make such a data acquisition and processing system a useful tool to perform noise analysis are: the improved characteristics of analog-to-digital converters employed to quantize the signals; the use of an on-line computer which allows a great accumulation and a rapid treatment of data together with an easy check of the correctness of the experiments; and the adoption of the time cross-correlation technique using two-detectors which by-pass the limitation of low efficiency detectors. (author)

UABUC - Single energy point model burnup computer code for water reactors

International Nuclear Information System (INIS)

El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

1981-01-01

UABUC is a single energy point reactor burnup computer program in FORTRAN language. The program calculates the change in the isotopic composition of the uranium fuel as a function of irradiation time with all its associated quantities such as the average point flux, the conversion ratio, macroscopic fuel cross sections, and the point reactivity profile. A step-wise time analytical solution was developed for the nonlinear first order burnup differential equations. The ''Westcott'' convention of the effective cross sections was used except for plutonium-240 and uranium-238. For plutonium-240, an effective microscopic cross section was derived from the direct physical arguments taking into account the selfshielding effect of plutonium-240 as well as the 1 ev. resonance absorption. For uranium-238, an effective cross section, reflecting the effect of fast fission and resonance absorption was used. The fission products were treated in the three groups with 50, 300, and 800 barns. The yields in the groups were treated as functions of the type of fissionable nuclides, the effective neutron temperature, and the epithermal index. Xenon-135 and Samarium-149 were treated separately as functions of irradiation time. (author)

Fusion reactor pumped laser

International Nuclear Information System (INIS)

Jassby, D.L.

1988-01-01

A nuclear pumped laser is described comprising: a toroidal fusion reactor, the reactor generating energetic neutrons; an annular gas cell disposed around the outer periphery of the reactor, the cell including an annular reflecting mirror disposed at the bottom of the cell and an annular output window disposed at the top of the cell; a gas lasing medium disposed within the annular cell for generating output laser radiation; neutron reflector material means disposed around the annular cell for reflecting neutrons incident thereon back into the gas cell; neutron moderator material means disposed between the reactor and the gas cell and between the gas cell and the neutron reflector material for moderating the energy of energetic neutrons from the reactor; converting means for converting energy from the moderated neutrons to energy pumping means for pumping the gas lasing medium; and beam compactor means for receiving output laser radiation from the annular output window and generating a single output laser beam therefrom

Voltage-Mode All-Pass Filters Including Minimum Component Count Circuits

Directory of Open Access Journals (Sweden)

Sudhanshu Maheshwari

2007-01-01

Full Text Available This paper presents two new first-order voltage-mode all-pass filters using a single-current differencing buffered amplifier and four passive components. Each circuit is compatible to a current-controlled current differencing buffered amplifier with only two passive elements, thus resulting in two more circuits, which employ a capacitor, a resistor, and an active element, thus using a minimum of active and passive component counts. The proposed circuits possess low output impedance, and hence can be easily cascaded for voltage-mode systems. PSPICE simulation results are given to confirm the theory.

Catalytic Reactor For Oxidizing Mercury Vapor

Science.gov (United States)

Helfritch, Dennis J.

1998-07-28

A catalytic reactor (10) for oxidizing elemental mercury contained in flue gas is provided. The catalyst reactor (10) comprises within a flue gas conduit a perforated corona discharge plate (30a, b) having a plurality of through openings (33) and a plurality of projecting corona discharge electrodes (31); a perforated electrode plate (40a, b, c) having a plurality of through openings (43) axially aligned with the through openings (33) of the perforated corona discharge plate (30a, b) displaced from and opposing the tips of the corona discharge electrodes (31); and a catalyst member (60a, b, c, d) overlaying that face of the perforated electrode plate (40a, b, c) opposing the tips of the corona discharge electrodes (31). A uniformly distributed corona discharge plasma (1000) is intermittently generated between the plurality of corona discharge electrode tips (31) and the catalyst member (60a, b, c, d) when a stream of flue gas is passed through the conduit. During those periods when corona discharge (1000) is not being generated, the catalyst molecules of the catalyst member (60a, b, c, d) adsorb mercury vapor contained in the passing flue gas. During those periods when corona discharge (1000) is being generated, ions and active radicals contained in the generated corona discharge plasma (1000) desorb the mercury from the catalyst molecules of the catalyst member (60a, b, c, d), oxidizing the mercury in virtually simultaneous manner. The desorption process regenerates and activates the catalyst member molecules.

Nanomechanical characterization by double-pass force-distance mapping

Energy Technology Data Exchange (ETDEWEB)

Dagdas, Yavuz S; Tekinay, Ayse B; Guler, Mustafa O; Dana, Aykutlu [UNAM Institute of Materials Science and Nanotechnology, Bilkent University, 06800 Ankara (Turkey); Necip Aslan, M, E-mail: aykutlu@unam.bilkent.edu.tr [Department of Physics, Istanbul Technical University, Istanbul (Turkey)

2011-07-22

We demonstrate high speed force-distance mapping using a double-pass scheme. The topography is measured in tapping mode in the first pass and this information is used in the second pass to move the tip over the sample. In the second pass, the cantilever dither signal is turned off and the sample is vibrated. Rapid (few kHz frequency) force-distance curves can be recorded with small peak interaction force, and can be processed into an image. Such a double-pass measurement eliminates the need for feedback during force-distance measurements. The method is demonstrated on self-assembled peptidic nanofibers.

Nanomechanical characterization by double-pass force-distance mapping

International Nuclear Information System (INIS)

Dagdas, Yavuz S; Tekinay, Ayse B; Guler, Mustafa O; Dana, Aykutlu; Necip Aslan, M

2011-01-01

We demonstrate high speed force-distance mapping using a double-pass scheme. The topography is measured in tapping mode in the first pass and this information is used in the second pass to move the tip over the sample. In the second pass, the cantilever dither signal is turned off and the sample is vibrated. Rapid (few kHz frequency) force-distance curves can be recorded with small peak interaction force, and can be processed into an image. Such a double-pass measurement eliminates the need for feedback during force-distance measurements. The method is demonstrated on self-assembled peptidic nanofibers.

Reactor perturbation calculations by Monte Carlo methods

International Nuclear Information System (INIS)

Gubbins, M.E.

1965-09-01

Whilst Monte Carlo methods are useful for reactor calculations involving complicated geometry, it is difficult to apply them to the calculation of perturbation worths because of the large amount of computing time needed to obtain good accuracy. Various ways of overcoming these difficulties are investigated in this report, with the problem of estimating absorbing control rod worths particularly in mind. As a basis for discussion a method of carrying out multigroup reactor calculations by Monte Carlo methods is described. Two methods of estimating a perturbation worth directly, without differencing two quantities of like magnitude, are examined closely but are passed over in favour of a third method based on a correlation technique. This correlation method is described, and demonstrated by a limited range of calculations for absorbing control rods in a fast reactor. In these calculations control rod worths of between 1% and 7% in reactivity are estimated to an accuracy better than 10% (3 standard errors) in about one hour's computing time on the English Electric KDF.9 digital computer. (author)

Spanish validation of the Premorbid Adjustment Scale (PAS-S).

Science.gov (United States)

Barajas, Ana; Ochoa, Susana; Baños, Iris; Dolz, Montse; Villalta-Gil, Victoria; Vilaplana, Miriam; Autonell, Jaume; Sánchez, Bernardo; Cervilla, Jorge A; Foix, Alexandrina; Obiols, Jordi E; Haro, Josep Maria; Usall, Judith

2013-02-01

The Premorbid Adjustment Scale (PAS) has been the most widely used scale to quantify premorbid status in schizophrenia, coming to be regarded as the gold standard of retrospective assessment instruments. To examine the psychometric properties of the Spanish version of the PAS (PAS-S). Retrospective study of 140 individuals experiencing a first episode of psychosis (n=77) and individuals who have schizophrenia (n=63), both adult and adolescent patients. Data were collected through a socio-demographic questionnaire and a battery of instruments which includes the following scales: PAS-S, PANSS, LSP, GAF and DAS-sv. The Cronbach's alpha was performed to assess the internal consistency of PAS-S. Pearson's correlations were performed to assess the convergent and discriminant validity. The Cronbach's alpha of the PAS-S scale was 0.85. The correlation between social PAS-S and total PAS-S was 0.85 (p<0.001); while for academic PAS-S and total PAS-S it was 0.53 (p<0.001). Significant correlations were observed between all the scores of each age period evaluated across the PAS-S scale, with a significance value less than 0.001. There was a relationship between negative symptoms and social PAS-S (0.20, p<0.05) and total PAS-S (0.22, p<0.05), but not with academic PAS-S. However, there was a correlation between academic PAS-S and general subscale of the PANSS (0.19, p<0.05). Social PAS-S was related to disability measures (DAS-sv); and academic PAS-S showed discriminant validity with most of the variables of social functioning. PAS-S did not show association with the total LSP scale (discriminant validity). The Spanish version of the Premorbid Adjustment Scale showed appropriate psychometric properties in patients experiencing a first episode of psychosis and who have a chronic evolution of the illness. Moreover, each domain of the PAS-S (social and academic premorbid functioning) showed a differential relationship to other characteristics such as psychotic symptoms, disability

A Study on the Flow Characterization in the Reactor Cavity

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jung; Ko, Kwang Jeok; Kim, Sung Hwan; Kim, Min Gyu; Cho, Yeon Ho; Kim, Hyun Min [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

2016-10-15

In this study, the flow characterization of the cooling air in reactor cavity nearby RCPSA has been analyzed by using a 3 dimensional model and the ANSYS CFX software in order to predict the Convective Heat Transfer Coefficient (CHTC) of the RCPSA. The Reactor Cavity is the annular space by the concrete structure, the Reactor Cavity Pool Seal Assembly (RCPSA), which consists of the welded steel and is designed to be installed between the RV and the refueling pool floor, and the Reactor Vessel (RV). For such reason, the RCPSA should be designed to provide the cooling air passage for ventilation to circulate high temperature air passing by the RV during the reactor operation. It means that the RCPSA is influenced by the convection of cooling air and the thermal expansion of the RV. Therefore, the flow characterization at the reactor cavity is one of the factors of the RCPSA design during the reactor operation. The flow distribution of the cooling air in reactor cavity nearby RCPSA has been analyzed using ANSYS CFX software to obtain the CHTC at surface of the RCPSA. 1) The temperature from the RV and the insulation is one of the critical factors for the thermal gradient of the cooling air and the CHTC in the reactor cavity. 2) The rapid change of the CHTC in inner region nearby inner and outer flexure is related to the geometry shape of the RCPSA and velocity of cooling air.

Instrument lance for boiling water reactors

International Nuclear Information System (INIS)

Proell, N.; Bertz, S.; Graebener, K.H.

1980-01-01

The instrument lance contains in the lance cover pipe a thimble as part of the drive chamber system. Other thimbles serve to carry neutron detectors. Detectors can be exchanged without opening the reactor pressure vessel and without removing the fuel elements. Furthermore the detector exchange is independent from the fuel element cycle. The measurement lance passes from the bottom of the pressure vessel over the total hight of the core in the water ducts between the fuel elements and can thus determine the neutron flux distribution. (DG) [de

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

Fission reactor recycling pump handling device

International Nuclear Information System (INIS)

Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru.

1991-01-01

This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.)

Fission reactor recycling pump handling device

Energy Technology Data Exchange (ETDEWEB)

Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru

1991-06-24

This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.).

Start-up analysis of INET-5 MW district heating prototype reactor

International Nuclear Information System (INIS)

Li Tianshu

1991-09-01

The main features and thermohydraulic design parameters of the INET-5 MW reactor (INET: Institute of Nuclear Technology of Tsinghua University, Beijing) are presented. The start-up process and the effect of thermohydraulic instability on start-up process have been analyzed. The main obstacle of start-up process of INET-5 MW reactor is to pass the instability region from 1 atm to normal operation condition. For avoiding instability, the start-up process should be divided into two steps. The results of three different start-up proposals calculated by DACOL code are given and compared. The possibility of instabilities for each proposal has been checked. The checked results show that there is no instability during start-up of the three proposals. So, it is supposed that the INET-5 MW reactor can safely and stably reach the operation conditions. Finally, some conclusions about the effect of instability on start-up in boiling mode of INET-5MW reactor are given

Emergency reactor core cooling facility

International Nuclear Information System (INIS)

Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka.

1996-01-01

The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

Emergency reactor core cooling facility

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

1996-11-01

The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

Conceptual Design of Angular Position Detector for Control Element Drive Mechanism of Small and Medium Reactor

International Nuclear Information System (INIS)

Yu, Je-Yong; Huh, Hyung; Kim, Ji-Ho; Choi, Suhn

2007-01-01

When the small and medium reactor is designed with a soluble boron free operation and nuclear heating for the reactor start-up, the design features require a Control Element Drive Mechanism (CEDM) to have a fine-step movement capability as well as a high reliability for a fine reactivity control. Also the reliability and accuracy of the information for the control rod position is important to the reactor safety as well as to design of the core protection system. The position signal of control rod is classified as a Class 1E because the rod position signal is used in the safety related systems. Therefore it will be separated from the control systems to the extent that a failure of any single control system component of a channel and shall have sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure. The position indicator is composed of a permanent magnet, reed switches and a voltage divider. Four independent position indicators around the upper pressure housing provide an indication of the position of a control rod comprising of a permanent magnet with a magnetic field concentrator which moves with the extension shaft connected to the control rod. The zigzag arranged reed switches are positioned along a line parallel to the path of the movement of the permanent magnet and it is activated selectively when the permanent magnet passes by. A voltage divider electrically connected to the reed switches provides a signal commensurate with the position of the control rod. The signal may then be transmitted to a position indicating device. But position indicator can not recognize the malfunction of the rotary step motor of CEDM instantly because its signal output is changed after the control rod moves more than a distance of reed switch interval

Culham conceptual Tokamak reactor MkII. Conceptual layout of buildings for a twin reactor power station

International Nuclear Information System (INIS)

Guthrie, J.A.S.; Harding, N.H.

1981-01-01

This paper discusses the conceptual design of the nuclear complex of a 2400 MWe twin fusion reactor power station utilising common services and a single containment building. The design is based upon environmental and maintenance logistical requirements, the provision of adequate storage, workshop and construction facilities and the constraints imposed by the geometry of the main and auxiliary reactor coolant systems. (author)

Balancing Radiation and Contrast Media Dose in Single-Pass Abdominal Multidetector CT: Prospective Evaluation of Image Quality.

Science.gov (United States)

Camera, Luigi; Romano, Federica; Liccardo, Immacolata; Liuzzi, Raffaele; Imbriaco, Massimo; Mainenti, Pier Paolo; Pizzuti, Laura Micol; Segreto, Sabrina; Maurea, Simone; Brunetti, Arturo

2015-11-01

As both contrast and radiation dose affect the quality of CT images, a constant image quality in abdominal contrast-enhanced multidetector computed tomography (CE-MDCT) could be obtained balancing radiation and contrast media dose according to the age of the patients. Seventy-two (38 Men; 34 women; aged 20-83 years) patients underwent a single-pass abdominal CE-MDCT. Patients were divided into three different age groups: A (20-44 years); B (45-65 years); and C (>65 years). For each group, a different noise index (NI) and contrast media dose (370 mgI/mL) was selected as follows: A (NI, 15; 2.5 mL/kg), B (NI, 12.5; 2 mL/kg), and C (NI, 10; 1.5 mL/kg). Radiation exposure was reported as dose-length product (DLP) in mGy × cm. For quantitative analysis, signal-to-noise (SNR) and contrast-to-noise (CNR) ratios were calculated for both the liver (L) and the abdominal aorta (A). Statistical analysis was performed with a one-way analysis of variance. Standard imaging criteria were used for qualitative analysis. Although peak hepatic enhancement was 152 ± 16, 128 ± 12, and 101 ± 14 Hounsfield units (P contrast media dose (mL) administered were 476 ± 147 and 155 ± 27 for group A, 926 ± 291 and 130 ± 16 for group B, and 1981 ± 451 and 106 ± 15 for group C, respectively (P contrast media dose administered to patients of different age. Copyright © 2015 AUR. Published by Elsevier Inc. All rights reserved.

BWR type reactors

International Nuclear Information System (INIS)

Tsunoyama, Shigeaki; Tanabe, Akira.

1979-01-01

Purpose: To provide a main steam pressure shock absorber for reflecting the effect of the pressure propagation to coolants surface in the reactor core. Constitution: An annular shock absorber having near the water level through holes for water level measurement is provided to the gap between the skirt of a steam separator and a pressure vessel. Pressure waves are made the rapid closure of a main steam check valve. If arrived from the dome to the shock absorber, are mostly reflected to the side of the dome and give no substantial effects on the water surface. If the through holes are made small enough, the effects of pressure waves passing through the holes are negligible if they reach the water surface. (Kawakami, Y.)

Reactor surface contamination stabilization. Innovative technology summary report

International Nuclear Information System (INIS)

1998-11-01

Contaminated surfaces, such as the face of a nuclear reactor, need to be stabilized (fixed) to avoid airborne contamination during decontamination and decommissioning activities, and to prepare for interim safe storage. The traditional (baseline) method of fixing the contamination has been to spray a coating on the surfaces, but ensuring complete coverage over complex shapes, such as nozzles and hoses, is difficult. The Hanford Site C Reactor Technology Demonstration Group demonstrated innovative technologies to assess stabilization properties of various coatings and to achieve complete coverage of complex surfaces on the reactor face. This demonstration was conducted in two phases: the first phase consisted of a series of laboratory assessments of various stabilization coatings on metal coupons. For the second phase, coatings that passed the laboratory tests were applied to the front face of the C Reactor and evaluated. The baseline coating (Rust-Oleum No. 769) and one of the innovative technologies did not completely cover nozzle assemblies on the reactor face, the most critical of the second-phase evaluation criteria. However, one of the innovative coating systems, consisting of a base layer of foam covered by an outer layer of a polymeric film, was successful. The baseline technology would cost approximately 33% as much as the innovative technology cost of $64,000 to stabilize an entire reactor face (196 m 2 or 2116 ft 2 ) with 2,004 nozzle assemblies, but the baseline system failed to provide complete surface coverage

Computational fluid dynamics simulations of light water reactor flows

International Nuclear Information System (INIS)

Tzanos, C.P.; Weber, D.P.

1999-01-01

Advances in computational fluid dynamics (CFD), turbulence simulation, and parallel computing have made feasible the development of three-dimensional (3-D) single-phase and two-phase flow CFD codes that can simulate fluid flow and heat transfer in realistic reactor geometries with significantly reduced reliance, especially in single phase, on empirical correlations. The objective of this work was to assess the predictive power and computational efficiency of a CFD code in the analysis of a challenging single-phase light water reactor problem, as well as to identify areas where further improvements are needed

Series lecture on advanced fusion reactors

International Nuclear Information System (INIS)

Dawson, J.M.

1983-01-01

The problems concerning fusion reactors are presented and discussed in this series lecture. At first, the D-T tokamak is explained. The breeding of tritium and the radioactive property of tritium are discussed. The hybrid reactor is explained as an example of the direct use of neutrons. Some advanced fuel reactions are proposed. It is necessary to make physics consideration for burning advanced fuel in reactors. The rate of energy production and the energy loss are important things. The bremsstrahlung radiation and impurity radiation are explained. The simple estimation of the synchrotron radiation was performed. The numerical results were compared with a more detailed calculation of Taimor, and the agreement was quite good. The calculation of ion and electron temperature was made. The idea to use the energy more efficiently is that one can take X-ray or neutrons, and pass them through a first wall of a reactor into a second region where they heat the material. A method to convert high temperature into useful energy is the third problem of this lecture. The device was invented by A. Hertzberg. The lifetime of the reactor depends on the efficiency of energy recovery. The idea of using spin polarized nuclei has come up. The spin polarization gives a chance to achieve a large multiplication factor. The advanced fuel which looks easiest to make go is D plus He-3. The idea of multipole is presented to reduce the magnetic field inside plasma, and discussed. Two other topics are explained. (Kato, T.)

A comparison of the radiological consequences of a HEU and LEU fueled research reactor

International Nuclear Information System (INIS)

Kollas, J.G.

1985-01-01

An analysis of the design basis accident radiological consequences of the HEU and LEU fueled Greek Research Reactor is presented. Doses and individual cancer risk from exposure to the passing radioactive cloud are estimated up to a distance of 20 km from the reactor site. Collective exposure and latent health effects are estimated for the total Athens area of 3081000 inhabitants. The results indicate that the plutonium isotopes buildup in the LEU fuel does not increase appreciably the consequences in respect to the HEU fueled reactor. The plutonium impact concerns mainly bone effects and secondly lung and whole body effects. The contribution to the limiting thyroid dose and the corresponding thyroid effects is insignificant. (author)

Neutronics calculations for denatured molten salt reactors: Assessing resource requirements and proliferation-risk attributes

International Nuclear Information System (INIS)

Ahmad, Ali; McClamrock, Edward B.; Glaser, Alexander

2015-01-01

Highlights: • We study the proliferation-risk and resource attributes of denatured MSRs. • MSRs offer significantly better resource efficiency compared to light-water reactors. • Denatured single-fluid MSRs reactors offer promising non-proliferation attributes. - Abstract: Molten salt reactors (MSRs) are often advocated as a radical but worthwhile alternative to traditional reactor concepts based on solid fuels. This article builds upon the existing research into MSRs to model and simulate the operation of thorium-fueled single-fluid and two-fluid reactors. The analysis is based on neutronics calculations and focuses on denatured MSR systems. Resource utilization and basic proliferation-risk attributes are compared to those of standard light-water reactors. Depending on specific design choices, even fully denatured reactors could reduce uranium and enrichment requirements by a factor of 3–4. Overall, denatured single-fluid designs appear as the most promising candidate technology minimizing both design complexity and overall proliferation risks despite being somewhat less attractive from the perspective of resource utilization

Influence of welding passes on grain orientation – The example of a multi-pass V-weld

International Nuclear Information System (INIS)

Ye, Jing; Moysan, Joseph; Song, Sung-Jin; Kim, Hak-Joon; Chassignole, Bertrand; Gueudré, Cécile; Dupond, Olivier

2012-01-01

The accurate modelling of grain orientations in a weld is important, when accurate ultrasonic test predictions of a welded assembly are needed. To achieve this objective, Electricité de France (EDF) and the Laboratoire de Caractérisation Non Destructive (LCND) have developed a dedicated code, which makes use of information recorded in the welding procedure. Among the welding parameters recorded, although the order in which the welding passes are made is of primary importance in the welding process, this information is not always well known or accurately described. In the present paper we analyse in greater detail the influence of the order of welding passes, using data obtained from the Centre for Advanced Non Destructive Evaluation (CANDE), derived from a dissimilar metal weld (DMW) with buttering. Comparisons are made using grain orientation measurements on a macrograph. - Highlights: ► Influence of welding process on grain structure is studied using the MINA model. ► For the first time the importance of a slight slope of the layers is evaluated. ► Two orders of passes are compared for the modelling approach. ► A major effect is observed due to a change in the order of passes.

Uranium, Plutonium and Neptunium Co-recovery with Irradiated Fast Reactor MOX Fuel by Single Cycle Extraction Process

Energy Technology Data Exchange (ETDEWEB)

Masaumi Nakahara; Yuichi Sano; Kazunori Nomura; Tadahiro Washiya; Jun Komaki [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)

2008-07-01

The behavior of Np in single cycle extraction processes using tri-n-butylphosphate (TBP) as an extractant for U, Pu and Np co-recovery was investigated as a part of NEXT (New Extraction System for Transuranium) process. Two approaches for Np co-recovery with U and Pu were carried out with irradiated MOX fuel from fast reactor 'JOYO'; one was the counter current experiment using a feed solution with a high HNO{sub 3} concentration and the other used a scrubbing solution with a high HNO{sub 3} concentration. Experimental results showed that the leakage of Np to the raffinate were 0.986 % and 5.96 % under the condition of high HNO{sub 3} concentration in the feed solution and scrubbing solution, respectively. The simulation results based on these experiments indicated that most of Np could be extracted and co-recovered with U and Pu, just by increasing HNO{sub 3} concentrations in the feed and scrubbing solution on the single cycle extraction process. (authors)

Enhancing load-following and/or spectral shift capability in single-sparger natural circulation boiling water reactors

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1992-01-01

This patent describes a method for obtaining load-following capability in a coiling water reactor (BWR) wherein housed within a reactor pressure vessel (RPV) is a nuclear core disposed within a shroud having a shroud head and which with the RPV defines an annulus region disposed beneath the nuclear core, an upper steam dome connected to a steam outlet in the RPV, a core upper plenum formed within the shroud head and disposed atop the nuclear core, a chimney mounted atop the shroud head and in fluid communication with the core upper plenum and with a steam separator having a skirt which is in fluid communication with the steam dome, the region outside of the chimney defining a downcomer region, there being a water level established therein under normal operation of the BWR, and the RPV containing a feedwater inlet. It comprises: disposing a single sparger connected to the feedwater inlet above the steam separator skirt bottom about the interior circumference of the RPV at an elevation at approximately the water level established during normal operation of the BWR; and adjusting the feedwater flow through the inlet and into the sparger to vary the water level to be above, at or below the elevational location of the sparger in response to load-following need

Neutronic design of a traveling wave reactor core

International Nuclear Information System (INIS)

Lopez S, R. C.; Francois L, J. L.

2010-10-01

The traveling wave reactor is an innovative kind of fast breeder reactor, capable of operate for decades without refueling and whose operation requires only a small amount of enriched fuel for the ignition. Also, one of its advantages is its versatility; it can be designed as small modules of about 100 M We or large scale units of 1000 M We. In this paper the behaviour of the traveling wave reactor core is studied in order to determine whether the traveling breeding/burning wave moves (as theoretically predicted) or not. To achieve this, we consider a two pieces cylinder, the first one, the ignition zone, containing highly enriched fuel and the second, the breeding zone, which is the larger, containing natural or depleted uranium or thorium. We consider that both zones are homogeneous mixtures of fuel, sodium as coolant and iron as structural material. We also include a reflector material outside the cylinder to reduce the neutron leakages. Simulations were run with MCNPX version 2.6 code. We observed that the wave does move as time passes as predicted by theory, and reactor remains supercritical in the time we have simulated (3000 days). Also, we found that thorium does not perform as well as uranium for breeding in this type of reactor. Further test with different reflectors are planned for both U-Pu and Th-U fuel cycles. (Author)

Convective cooling in a pool-type research reactor

Science.gov (United States)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

Deformational Features and Microstructure Evolution of Copper Fabricated by a Single Pass of the Elliptical Cross-Section Spiral Equal-Channel Extrusion (ECSEE) Process

Science.gov (United States)

Wang, Chengpeng; Li, Fuguo; Liu, Juncheng

2018-04-01

The objectives of this work are to study the deformational feature, textures, microstructures, and dislocation configurations of ultrafine-grained copper processed by the process of elliptical cross-section spiral equal-channel extrusion (ECSEE). The deformation patterns of simple shear and pure shear in the ECSEE process were evaluated with the analytical method of geometric strain. The influence of the main technical parameters of ECSEE die on the effective strain distribution on the surface of ECSEE-fabricated samples was examined by the finite element simulation. The high friction factor could improve the effective strain accumulation of material deformation. Moreover, the pure copper sample fabricated by ECSEE ion shows a strong rotated cube shear texture. The refining mechanism of the dislocation deformation is dominant in copper processed by a single pass of ECSEE. The inhomogeneity of the micro-hardness distribution on the longitudinal section of the ECSEE-fabricated sample is consistent with the strain and microstructure distribution features.

Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment

KAUST Repository

Zhu, Xiuping

2013-05-01

Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H2O2 generation and Fe2+ release for the Fenton reaction. In tests using phenol, 75±2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants. © 2013 Elsevier B.V.

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

International Nuclear Information System (INIS)

Chenu, A.

2011-10-01

Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

Passing the baton

CERN Multimedia

2011-01-01

It was not only in South Korea that batons were being passed last week. While the cream of the world’s athletes were competing in the World Athletics Championships, the cream of the world’s accelerator scientists were on their way to San Sebastian in Spain for the International Particle Accelerator Conference.  One of them was carrying a rather special baton for a handover of a different kind.   When Fermilab’s Vladimir Shiltsev handed the high-energy frontier baton to CERN’s Mike Lamont on Tuesday, it marked the end of an era: a time to look back on the phenomenal contribution the Tevatron has made to particle physics over its 25-year operational lifetime, and the great contribution Fermilab has made over that period to global collaboration in particle physics. There’s always a lot of emotion involved in passing the baton. In athletics, it’s the triumph of wining or the heartbreak of losing. But for this special baton, the...

Fish pass assessment by remote control: a novel framework for quantifying the hydraulics at fish pass entrances

Science.gov (United States)

Kriechbaumer, Thomas; Blackburn, Kim; Gill, Andrew; Breckon, Toby; Everard, Nick; Wright, Ros; Rivas Casado, Monica

2014-05-01

Fragmentation of aquatic habitats can lead to the extinction of migratory fish species with severe negative consequences at the ecosystem level and thus opposes the target of good ecological status of rivers defined in the EU Water Framework Directive (WFD). In the UK, the implementation of the EU WFD requires investments in fish pass facilities of estimated 532 million GBP (i.e. 639 million Euros) until 2027 to ensure fish passage at around 3,000 barriers considered critical. Hundreds of passes have been installed in the past. However, monitoring studies of fish passes around the world indicate that on average less than half of the fish attempting to pass such facilities are actually successful. There is a need for frameworks that allow the rapid identification of facilities that are biologically effective and those that require enhancement. Although there are many environmental characteristics that can affect fish passage success, past research suggests that variations in hydrodynamic conditions, reflected in water velocities, velocity gradients and turbulences, are the major cues that fish use to seek migration pathways in rivers. This paper presents the first steps taken in the development of a framework for the rapid field-based quantification of the hydraulic conditions downstream of fish passes and the assessment of the attractivity of fish passes for salmonids and coarse fish in UK rivers. For this purpose, a small-sized remote control platform carrying an acoustic Doppler current profiler (ADCP), a GPS unit, a stereo camera and an inertial measurement unit has been developed. The large amount of data on water velocities and depths measured by the ADCP within relatively short time is used to quantify the spatial and temporal distribution of water velocities. By matching these hydraulic features with known preferences of migratory fish, it is attempted to identify likely migration routes and aggregation areas at barriers as well as hydraulic features that

Decommissioning of the CANDU-PHW reactor

International Nuclear Information System (INIS)

Unsworth, G.N.

1977-04-01

This report contains the results of a study of various aspects of decommissioning of reactors. The study places in perspective the size of the job, the hazards involved, the cost and the environmental impact. The three internationally agreed ''stages'' of decommissioning, namely, mothballing, entombment, and dismantling are defined and discussed. The single unit 600 MW(e) CANDU is chosen as the type of reactor on which the discussion is focussed but the conclusions reached will provide a basis for judgement of the costs and problems associated with decommissioning reactors of other sizes and types. (author)

Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

International Nuclear Information System (INIS)

Bays, Samuel; Medvedev, Pavel; Pope, Michael; Ferrer, Rodolfo; Forget, Benoit; Asgari, Mehdi

2009-01-01

This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

Tandem Mirror Reactor Systems Code (Version I)

International Nuclear Information System (INIS)

Reid, R.L.; Finn, P.A.; Gohar, M.Y.

1985-09-01

A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

Roof slab cooling device in a FBR type reactor

International Nuclear Information System (INIS)

Tarutani, Kohei

1987-01-01

Purpose: To obtain a roof slab cooling device capable of retaining cooling performance even in a case of electric power supply stop or failure and effective from economical point of view. Constitution: Atmospheric air is introduced into the cooling chamber of a proof slab and spontaneously passed to a exit pipeway connected to a stack thereby cooling the roof slab. Specifically, atmospheric air entered from the inlet pipeway is introduced to the cooling chamber and absorbs heat generate from the inside of the reactor container. Warmed air is sucked from the exit pipeway and then released into the atmosphere passing through the stack. The air cools the roof slab during circulation due to spontaneous passage and keeps the slab at a low temperature. Since the air is passed spontaneously, no power such as for a blower is required at all and, if the electric power supply should be lost, the cooling power can be maintained as it is to provide a high reliability. Further, since no electric power is required for the blowing power, it has high economical merit. (Horiuchi, T.)

Fast reactor programme

International Nuclear Information System (INIS)

Plakman, J.C.

1982-01-01

This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

Breeding description for fast reactors and symbiotic reactor systems

International Nuclear Information System (INIS)

Hanan, N.A.

1979-01-01

A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

A system for regulating the pressure of resuperheated steam in high temperature gas-cooled reactor power stations

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegines, K.O.

1975-01-01

The invention relates to a system for regulating steam-pressure in the re-superheating portion of a steam-boiler receiving heat from a gas-cooled high temperature nuclear reactor, provided with gas distributing pumps driven by steam-turbines. The system comprises means for generating a pressure signal of desired magnitude for the re-superheating portion, and means for providing a real pressure in the re-superheating portion, means (including a by-passing device) for generating steam-flow rate signal of desired magnitude, a turbine by-pass device comprising a by-pass tapping means for regulating the steam-flow-rate in said turbine according to the desired steam-flow rate signal and means for controlling said by-pass tapping means according to said desired steam-flow-rate signal [fr

Flow visualization study of two phase flow in a single bend outlet feeder pipe and horizontal annulus of outlet end-fitting of a CANDU reactor

International Nuclear Information System (INIS)

Supa-Amornkul, S.; Lister, D.H.; Steward, F.R.

2005-01-01

'Full text:' In CANDU-6 reactors, the pressurized high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310 o C with up to 0.30 steam voidage, turns through 90 o as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations have reported excessive corrosion of their outlet feeder pipes; especially between the first metre, which consisted of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow accelerated corrosion. In order to understand the hydrodynamics of the coolant in the outlet feeders by flow-visualization, a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting were fabricated. The feeder consisted of a 54 mm inside diameter acrylic pipe with a 73 o bend, connecting to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outer diameter, and an outer pipe, 150 mm inner diameter, both 190.7 cm long in length. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 19 L/s and the volume fraction of air varied from 0.05 to 0.56. The phase distributions within the feeder pipe and along the length of the annulus were investigated with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Particular attention was paid to the flow pattern at the inside of the bend, where a CFD

Flow visualization study of two phase flow in a single bend outlet feeder pipe and horizontal annulus of outlet end-fitting of a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Supa-Amornkul, S.; Lister, D.H.; Steward, F.R. [Univ. of New Brunswick, Fredericton, New Brunswick (Canada)]. E-mail: h796e@unb.ca; dlister@unb.ca; fsteward@unb.ca

2005-07-01

'Full text:' In CANDU-6 reactors, the pressurized high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310{sup o}C with up to 0.30 steam voidage, turns through 90{sup o} as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations have reported excessive corrosion of their outlet feeder pipes; especially between the first metre, which consisted of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow accelerated corrosion. In order to understand the hydrodynamics of the coolant in the outlet feeders by flow-visualization, a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting were fabricated. The feeder consisted of a 54 mm inside diameter acrylic pipe with a 73{sup o} bend, connecting to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outer diameter, and an outer pipe, 150 mm inner diameter, both 190.7 cm long in length. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 19 L/s and the volume fraction of air varied from 0.05 to 0.56. The phase distributions within the feeder pipe and along the length of the annulus were investigated with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Particular attention was paid to the flow pattern at the inside

Water Gas Shift Reaction with A Single Stage Low Temperature Membrane Reactor

Energy Technology Data Exchange (ETDEWEB)

Ciora, Richard J [Media and Process Technology Inc., Pittsburgh, PA (United States); Liu, Paul KT [Media and Process Technology Inc., Pittsburgh, PA (United States)

2013-12-31

Palladium membrane and Palladium membrane reactor were developed under this project for hydrogen separation and purification for fuel cell applications. A full-scale membrane reactor was designed, constructed and evaluated for the reformate produced from a commercial scale methanol reformer. In addition, the Pd membrane and module developed from this project was successfully evaluated in the field for hydrogen purification for commercial fuel cell applications.

Reactor surveillance by noise analysis

International Nuclear Information System (INIS)

Ciftcioglu, Ozer

1988-01-01

A real-time noise analysis system is designed for the TRIGA reactor at Istanbul Technical University. By means of the noise techniques, reactor surveillance is performed together with failure diagnosis. The fast data processing is carried out by FFT in real-time so that malfunction or non-stationary operation of the reactor in long term can be identified by comparing the noise power spectra with the corresponding reference patterns while the decision making procedure is accomplished by the method of hypothesis testing. The system being computer based safety instrumentation involves CAMAC in conjunction with the RT-11 (PDP-11) single user dedicated environment. (author)

Analysis of Biomechanical Structure and Passing Techniques in Basketball

OpenAIRE

Ricardo E. Izzo; Luca Russo

2011-01-01

The basketball is a complex sport, which these days has become increasingly linked to its’ psychophysical aspects rather than to the technical ones. Therefore, it is important to make a through study of the passing techniques from the point of view of the type of the pass and its’ biomechanics. From the point of view of the type of the used passes, the most used is the two-handed chest pass with a frequency of 39.9%. This is followed, in terms of frequency, by one-handed passes – the baseball...

Modelling of sludge blanket height and flow pattern in UASB reactors treating municipal wastewater

International Nuclear Information System (INIS)

Singh, K.S.; Viraraghavan, T.

2002-01-01

Two upflow anaerobic sludge blanket (UASB) reactors were started-up and operated for approximately 900 days to examine the feasibility of treating municipal wastewater under low temperature conditions. A modified solid distribution model was formulated by incorporating the variation of biogas production rate with a change in temperature. This model was used to optimize the sludge blanket height of UASB reactors for an effective operation of gas-liquid-solid (GLS) separation device. This model was found to simulate well the solid distribution as confirmed experimental observation of solid profile along the height of the reactor. Mathematical analysis of tracer curves indicated the presence of a mixed type of flow pattern in the sludge-bed zone of the reactor. It was found that the dead-zone and by-pass flow fraction were impacted by the change in operating temperatures. (author)

Methodology of the On-Iine FoIIow Simulation of Pebble-bed High-temperature Reactors

International Nuclear Information System (INIS)

Xia Bing; Li Fu; Wei Chunlin; Zheng Yanhua; Chen Fubing; Zhang Jian; Guo Jiong

2014-01-01

The on-line fuel management is an essential feature of the pebble-bed high-temperature reactors (PB-HTRs), which is strongly coupled with the normal operation of the reactor. For the purpose of on-line analysis of the continuous shuffling scheme of numerous fuel pebbles, the follow simulation upon the real operation is necessary for the PB-HTRs. In this work, the on-line follow simulation methodology of the PB-HTRs’ operation is described, featured by the parallel treatments of both neutronics analysis and fuel cycling simulation. During the simulation, the operation history of the reactor is divided into a series of burn-up cycles according to the behavior of operation data, in which the steady-state neutron transport equations are solved and the diffusion theory is utilized to determine the physical features of the reactor core. The burn-up equations of heavy metals, fission products and neutron poisons including B-10, decoupled from the pebble flow term, are solved to analyze the burn-up process within a single burn-up cycle. The effect of pebble flow is simulated separately through a discrete fuel shuffling pattern confined by curved pebble flow channels, and the effect of multiple pass of the fuel is represented by logical batches within each spatial region of the core. The on-line thermal-hydraulics feedback is implemented for each bur-up cycle by using the real thermal-hydraulics data of the core operation. The treatment of control rods and absorber balls is carried out by utilizing a coupled neutron transport-diffusion calculation along with discontinuity factors. The physical models mentioned above are established mainly by using a revised version of the V.S.O.P program system. The real operation data of HTR-10 is utilized to verify the methodology presented in this work, which gives good agreement between simulation results and operation data. (author)

System modeling for the advanced thermionic initiative single cell thermionic space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Lewis, B.R.; Klein, A.C.; Pawlowski, R.A.

1993-01-01

Incore thermionic space reactor design concepts which operate in a nominal power output range of 20 to 40 kWe are described. Details of the neutronics, thermionic, shielding, and heat rejection performance are presented. Two different designs, ATI-Driven and ATI-Driverless, are considered. Comparison of the core overall performance of these two configurations are described. The comparison of these two cores includes the overall conversion efficiency, reactor mass, shield mass, and heat rejection mass. An overall system design has been developed to model the advanced incore thermionic energy conversion based nuclear reactor systems for space applications in this power range

A Voltage Gain-Controlled Modified CFOA And Its Application in Electronically Tunable Four-Mode All-Pass Filter Design

OpenAIRE

Norbert Herencsar; Jaroslav Koton; Abhirup Lahiri; Bilgin Metin; Kamil Vrba

2012-01-01

This paper presents a new active building block (ABB) called voltage gain-controlled modified current feedback amplifier (VGC-MCFOA) based on bipolar junction transistor technology. The versatility of the new ABB is demonstrated in new first-order all-pass filter structure design employing single VGC-MCFOA, single grounded capacitor, and three resistors. Introduced circuit provides all four possible transfer functions at the same configuration, namely current-mode, transimpedance-mode, transa...

Report on generation IV technical working group 3 : liquid metal reactors

International Nuclear Information System (INIS)

Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

2002-01-01

This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

On application of CFD codes to problems of nuclear reactor safety

International Nuclear Information System (INIS)

Muehlbauer, Petr

2005-01-01

The 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in May 2002 at Aix-en-Province, France, recommended formation of writing groups to report the need of guidelines for use and assessment of CFD in single-phase nuclear reactor safety problems, and on recommended extensions to CFD codes to meet the needs of two-phase problems in nuclear reactor safety. This recommendations was supported also by Working Group on the Analysis and Management of Accidents and led to formation oaf three Writing Groups. The first writing Group prepared a summary of existing best practice guidelines for single phase CFD analysis and made a recommendation on the need for nuclear reactor safety specific guidelines. The second Writing Group selected those nuclear reactor safety applications for which understanding requires or is significantly enhanced by single-phase CFD analysis, and proposed a methodology for establishing assesment matrices relevant to nuclear reactor safety applications. The third writing group performed a classification of nuclear reactor safety problems where extension of CFD to two-phase flow may bring real benefit, a classification of different modeling approaches, and specification and analysis of needs in terms of physical and numerical assessments. This presentation provides a review of these activities with the most important conclusions and recommendations (Authors)

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

Science.gov (United States)

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

Optimization of a radially cooled pebble bed reactor - HTR2008-58117

International Nuclear Information System (INIS)

Boer, B.; Kloosterman, J. L.; Lathouwers, D.; Van Der Hagen, T. H. J. J.; Van Dam, H.

2008-01-01

By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained. The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal. The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as I cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling. Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop (Δp = -2.6 bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (ΔT = -50 deg. C) can be achieved

Fission reactor control rod

International Nuclear Information System (INIS)

Irie, Tomoo.

1991-01-01

The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

Present state of the liner of the reactor

International Nuclear Information System (INIS)

Aguilar H, F.; Raya A, R.; Mazon R, R.

2001-07-01

When being presented to work the operation personnel of the reactor, on Monday January 10, 1983, they noticed that the reactor pool was overflowing of water and the floor of the room was partially flooded. The personnel proceeded to revise the feedwater systems to the pool, the Emergency Cooling System of the core and that of Water of Reinstatement, was found that the passing valve of this last it was lightly open. It was discovered that the water that was flooded in the floor of the room it came from the relief valves of the ports TW-1 and RW-2 and of three glides that were in the Thermal Column area. It was proceeded to lower the one level of water of the pool to their normal position and it was clean the water flooded in the salts. (Author)

PRISM: An innovative liquid metal fast breeder reactor

International Nuclear Information System (INIS)

Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

1986-01-01

This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design

Retail Bank Interest Rate Pass-Through; Is Chile Atypical?

OpenAIRE

Alessandro Rebucci; Marco A Espinosa-Vega

2003-01-01

This paper investigates empirically the pass-through of money market interest rates to retail banking interest rates in Chile, the United States, Canada, Australia, New Zealand, and five European countries. Overall, Chile's pass-through does not appear atypical. Based on a standard error-correction model, we find that, as in most countries considered, Chile's measured pass-through is incomplete. But Chile's pass-through is also faster than in many other countries considered and is comparable ...

North Texas Sediment Budget: Sabine Pass to San Luis Pass

Science.gov (United States)

2006-09-01

concrete units have been placed over sand-filled fabric tube . .......................................33 Figure 28. Sand-filled fabric tubes protecting...system UTM Zone 15, NAD 83 Longshore drift directions King (in preparation) Based on wave hindcast statistics and limited buoy data Rollover Pass...along with descriptions of the jetties and limited geographic coordinate data1 (Figure 18). The original velum or Mylar sheets from which the report

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

Science.gov (United States)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control

Differentiating benign and malignant breast lesions with T2*-weighted first pass perfusion imaging

International Nuclear Information System (INIS)

Kvistad, K.A.; Smenes, E.; Haraldseth, O.; Lundgren, S.; Fjoesne, H.E.; Smethurst, H.B.

1999-01-01

Purpose: Invasive breast carcinomas and fibroadenomas are often difficult to differentiate in dynamic contrast-enhanced T1-weighted MR imaging of the breast, because both tumors can enhance strongly after contrast injection. The purpose of this study was to evaluate whether the addition of T2*-weighted first pass perfusion imaging can increase the differentiation of malignant from benign lesions. Material and Methods: Nine patients with invasive carcinomas and 10 patients with contrast enhancing fibroadenomas were examined by a dynamic contrast-enhanced T1-weighted 3D sequence immediately followed by a single slice T2*-weighted first pass perfusion sequence positioned in the contrast-enhancing lesion. Results: The carcinomas and the fibroadenomas were impossible to differentiate based on the contrast enhancement characteristics in the T1-weighted sequence. The signal loss in the T2*-weighted perfusion sequence was significantly stronger in the carcinomas than in the fibroadenomas (p=0.0004). Conclusion: Addition of a T2*-weighted first pass perfusion sequence with a high temporal resolution can probably increase the differentiation of fibroadenomas from invasive carcinomas in contrast-enhanced MR imaging of the breast. (orig.)

Development of an advanced antineutrino detector for reactor monitoring

Energy Technology Data Exchange (ETDEWEB)

Classen, T., E-mail: classen2@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bernstein, A.; Bowden, N.S. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Cabrera-Palmer, B. [Sandia Livermore National Laboratories, Livermore, CA 94550 (United States); Ho, A.; Jonkmans, G. [Atomic Energy of Canada, Limited, Chalk River Laboratories, Chalk River, ON (Canada); Kogler, L.; Reyna, D. [Sandia Livermore National Laboratories, Livermore, CA 94550 (United States); Sur, B. [Atomic Energy of Canada, Limited, Chalk River Laboratories, Chalk River, ON (Canada)

2015-01-21

Here we present the development of a compact antineutrino detector for the purpose of nuclear reactor monitoring, improving upon a previously successful design. This paper will describe the design improvements of the detector which increases the antineutrino detection efficiency threefold over the previous effort. There are two main design improvements over previous generations of detectors for nuclear reactor monitoring: dual-ended optical readout and single volume detection mass. The dual-ended optical readout eliminates the need for fiducialization and increases the uniformity of the detector's optical response. The containment of the detection mass in a single active volume provides more target mass per detector footprint, a key design criteria for operating within a nuclear power plant. This technology could allow for real-time monitoring of the evolution of a nuclear reactor core, independent of reactor operator declarations of fuel inventories, and may be of interest to the safeguards community.

Segmental-dependent membrane permeability along the intestine following oral drug administration: Evaluation of a triple single-pass intestinal perfusion (TSPIP) approach in the rat.

Science.gov (United States)

Dahan, Arik; West, Brady T; Amidon, Gordon L

2009-02-15

In this paper we evaluate a modified approach to the traditional single-pass intestinal perfusion (SPIP) rat model in investigating segmental-dependent permeability along the intestine following oral drug administration. Whereas in the traditional model one single segment of the intestine is perfused, we have simultaneously perfused three individual segments of each rat intestine: proximal jejunum, mid-small intestine and distal ileum, enabling to obtain tripled data from each rat compared to the traditional model. Three drugs, with different permeabilities, were utilized to evaluate the model: metoprolol, propranolol and cimetidine. Data was evaluated in comparison to the traditional method. Metoprolol and propranolol showed similar P(eff) values in the modified model in all segments. Segmental-dependent permeability was obtained for cimetidine, with lower P(eff) in the distal parts. Similar P(eff) values for all drugs were obtained in the traditional method, illustrating that the modified model is as accurate as the traditional, throughout a wide range of permeability characteristics, whether the permeability is constant or segment-dependent along the intestine. Three-fold higher statistical power to detect segmental-dependency was obtained in the modified approach, as each subject serves as his own control. In conclusion, the Triple SPIP model can reduce the number of animals utilized in segmental-dependent permeability research without compromising the quality of the data obtained.

Convective cooling in a pool-type research reactor

Energy Technology Data Exchange (ETDEWEB)

Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

Energy Technology Data Exchange (ETDEWEB)

Liu, Hongbing, E-mail: liuhb07@mails.tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Du, Dong, E-mail: dudong@tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Huang, An; Chang, Baohua; Han, Zandong [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); He, Ayada [Shanghai Electric Power Generation Group Shanghai Generator Works, Shanghai 200240 (China)

2016-08-15

Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

International Nuclear Information System (INIS)

Liu, Hongbing; Du, Dong; Huang, An; Chang, Baohua; Han, Zandong; He, Ayada

2016-01-01

Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Vikhorev, Yu.V.; Biryukov, G.I.; Kirilyuk, N.A.; Lobanov, V.N.

1977-01-01

A fuel assembly is proposed for nuclear reactors allowing remote replacement of control rod bundles or their shifting from one assembly to another, i.e., their multipurpose use. This leads to a significant increase in fuel assembly usability. In the fuel assembly the control rod bundle is placed in guide tube channels to which baffles are attached for fuel element spacing. The remote handling of control rods is provided by a hollow cylinder with openings in its lower bottom through which the control rods pass. All control rods in a bundle are mounted to a cross beam which in turn is mounted in the cylinder and is designed for grasping the whole rod bundle by a remotely controlled telescopic mechanism in bundle replacement or shifting. (Z.M.)

Analysis of Biomechanical Structure and Passing Techniques in Basketball

Directory of Open Access Journals (Sweden)

Ricardo E. Izzo

2011-06-01

Full Text Available The basketball is a complex sport, which these days has become increasingly linked to its’ psychophysical aspects rather than to the technical ones. Therefore, it is important to make a through study of the passing techniques from the point of view of the type of the pass and its’ biomechanics. From the point of view of the type of the used passes, the most used is the two-handed chest pass with a frequency of 39.9%. This is followed, in terms of frequency, by one-handed passes – the baseball, with 20.9 % – and by the two-handed over the head pass, with 18.2 %, and finally, one- or two-handed indirect passes (bounces, with 11.2 % and 9.8 %. Considering the most used pass in basketball, from the biomechanical point of view, the muscles involved in the correct movement consider all the muscles of the upper extremity, adding also the shoulder muscles as well as the body fixators (abdominals, hip flexors, knee extensors, and dorsal flexors of the foot. The technical and conditional analysis considers the throwing speed, the throw height and the air resistance. In conclusion, the aim of this study is to give some guidelines to improve the mechanical execution of the movements in training, without neglecting the importance of the harmony of the movements themselves.

Mouse myocardial first-pass perfusion MR imaging

NARCIS (Netherlands)

Coolen, Bram F.; Moonen, Rik P. M.; Paulis, Leonie E. M.; Geelen, Tessa; Nicolay, Klaas; Strijkers, Gustav J.

2010-01-01

A first-pass myocardial perfusion sequence for mouse cardiac MRI is presented. A segmented ECG-triggered acquisition combined with parallel imaging acceleration was used to capture the first pass of a Gd-DTPA bolus through the mouse heart with a temporal resolution of 300-400 msec. The method was

Reactor protection system design using micro-computers

International Nuclear Information System (INIS)

Fairbrother, D.B.

1977-01-01

Reactor Protection Systems for Nuclear Power Plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. This paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

Reactor protection system design using micro-computers

International Nuclear Information System (INIS)

Fairbrother, D.B.

1976-01-01

Reactor protection systems for nuclear power plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. The paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

Performance of a UASB reactor treating coffee wet wastewater

International Nuclear Information System (INIS)

Guardia Puebla, Yans; Rodríguez Pérez, Suyén; Janet Jiménez Hernández; Sánchez Girón, Víctor

2014-01-01

The present work shows the results obtained in the anaerobic digestion process of coffee wet wastewater processing. An UASB anaerobic reactor was operated in single-stage in mesophilic temperature controlled conditions (37±1ºC). The effect of both organic loading rate (OLR) and hydraulic retention time (HRT) in the anaerobic digestion of coffee wet wastewater was investigated. The OLR values considered in the single-stage UASB reactor varied in a range of 3,6-4,1 kgCOD m-3 d-1 and the HRT stayed in a range of 21,5-15,5 hours. The evaluation results show that the best performance of UASB reactor in single-stage was obtained at OLR of 3,6 kg COD m-3 d-1 with an average value of total and soluble COD removal of 77,2% and 83,4%, respectively, and average methane concentration in biogas of 61%. The present study suggests that the anaerobic digestion is suitable to treating coffee wet wastewater. (author)

Mouse myocardial first-pass perfusion MR imaging

NARCIS (Netherlands)

Coolen, B.F.; Moonen, R.P.M.; Paulis, L.E.M.; Geelen, T.; Nicolay, K.; Strijkers, G.J.

2010-01-01

A first-pass myocardial perfusion sequence for mouse cardiac MRI is presented. A segmented ECG-triggered acquisition combined with parallel imaging acceleration was used to capture the first pass of a Gd-DTPA bolus through the mouse heart with a temporal resolution of 300–400 msec. The method was

12 CFR 560.32 - Pass-through investments.

Science.gov (United States)

2010-01-01

... 12 Banks and Banking 5 2010-01-01 2010-01-01 false Pass-through investments. 560.32 Section 560.32 Banks and Banking OFFICE OF THRIFT SUPERVISION, DEPARTMENT OF THE TREASURY LENDING AND INVESTMENT Lending and Investment Powers for Federal Savings Associations § 560.32 Pass-through investments. (a) A...

Single-pass, efficient type-I phase-matched frequency doubling of high-power ultrashort-pulse Yb-fiber laser using LiB_3O_5

Science.gov (United States)

Shukla, Mukesh Kumar; Kumar, Samir; Das, Ritwick

2016-05-01

We report 48 % efficient single-pass second harmonic generation of high-power ultrashort-pulse ({≈ }250 fs) Yb-fiber laser by utilizing type-I phase matching in LiB_3O_5 (LBO) crystal. The choice of LBO among other borate crystals for high-power frequency doubling is essentially motivated by large thermal conductivity, low birefringence and weak group velocity dispersion. By optimally focussing the beam in a 4-mm-long LBO crystal, we have generated about 2.3 W of average power at 532 nm using 4.8 W of available pump power at 1064 nm. The ultrashort green pulses were found out to be near-transform limited sech^2 pulses with a pulse width of Δ τ ≈ 150 fs and being delivered at 78 MHz repetition rate. Due to appreciably low spatial walk-off angle for LBO ({≈ }0.4°), we obtain M^2beam which signifies marginal distortion in comparison with the pump beam (M^2<1.15). We also discuss the impact of third-order optical nonlinearity of the LBO crystal on the generated ultrashort SH pulses.

Coolant cleanup system for BWR type reactor

International Nuclear Information System (INIS)

Kinoshita, Shoichiro; Araki, Hidefumi.

1993-01-01

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

Pass-Through to Import Prices: Evidence from Developing Countries

OpenAIRE

Miguel Fuentes

2007-01-01

In this paper I study the pass-through of nominal exchange rate changes to the price of imported goods in four developing countries. The results indicate that 75% of changes in the exchange rate are passed-through to the domestic currency price of imported goods within one quarter. Complete pass-through is attained within one year. There is no evidence that exchange rate pass-through to the price of imported goods has declined over time even in those countries that have managed to reduce infl...

Liquid metal cooled nuclear reactor

International Nuclear Information System (INIS)

Guidez, Joel; Jarriand, Paul.

1975-01-01

The invention concerns a fast neutron nuclear reactor cooled by a liquid metal driven through by a primary pump of the vertical drive shaft type fitted at its lower end with a blade wheel. To each pump is associated an exchanger, annular in shape, fitted with a central bore through which passes the vertical drive shaft of the pump, its wheel being mounted under the exchanger. A collector placed under the wheel comprises an open upward suction bell for the liquid metal. A hydrostatic bearing is located above the wheel to guide the drive shaft and a non detachable diffuser into which at least one delivery pipe gives, envelopes the wheel [fr

36 CFR 13.918 - Sable Pass Wildlife Viewing Area.

Science.gov (United States)

2010-07-01

... 36 Parks, Forests, and Public Property 1 2010-07-01 2010-07-01 false Sable Pass Wildlife Viewing... Preserve General Provisions § 13.918 Sable Pass Wildlife Viewing Area. (a) Entry into the Sable Pass Wildlife Viewing Area is prohibited from May 1 to September 30 unless authorized by the Superintendent. (b...

Time-varying exchange rate pass-through: experiences of some industrial countries

OpenAIRE

Toshitaka Sekine

2006-01-01

This paper estimates exchange rate pass-through of six major industrial countries using a time-varying parameter with stochastic volatility model. Exchange rate pass-through is divided into impacts of exchange rate fluctuations to import prices (first-stage pass-through) and those of import price movements to consumer prices (second-stage pass-through). The paper finds that both stages of pass-through have declined over time for all the sample countries. The decline in second-stage pass-throu...

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

Science.gov (United States)

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

International Nuclear Information System (INIS)

John Darrell Bess

2008-01-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

Analysis of pressure drop accidents in high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kameoka, Toshiyuki

1980-01-01

Research and development are carried out on various problems in order to realize a multi-purpose, high temperature gas-cooled experimental reactor by Japan Atomic Energy Research Institute and others. In the experimental reactor in consideration at present, it is planned to flow helium at 1000 deg C and 40 atm. For the purpose, high temperature heat insulation structures are designed and developed, which insulate heat on the internal surfaces of pressure vessels and pipings. Consideration must be given to these internal heat insulation structures about the various characteristics in the working environmental temperature and pressure conditions, the measures for preventing the by-pass flow due to the formation of gaps and the abnormal leak of heat through the natural convection in the heat insulators and others. In this paper, the experimental results on the rapid pressure reduction characteristics of ceramic fiber heat insulation structures are reported. The ceramic fiber heat insulation structures have the features such as the application to uneven surfaces and penetration parts, the prevention of by-pass flow, and very low permeability. The problem is the restoring force after the high temperature compression. The experiment on rapid pressure reduction due to the accidental release of gas and the results are reported. (Kako, I.)

Instrumentation amplifier implements second-order active low-pass filter with high gain factor

International Nuclear Information System (INIS)

Blomqvist, Kim H; Eskelinen, Pekka; Sepponen, Raimo E

2011-01-01

A single-ended second-order active low-pass filter can simultaneously provide high gain factor and dc voltage subtraction. This makes it possible to reduce the number of components and signal processing stages needed in an application where small voltage changes are measured on the top of large dc voltage masked by a large amplitude oscillating carrier. The filter described in this paper is constructed from a conventional 3-op-amp instrumentation amplifier and five passive circuit elements. (technical design note)

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

Science.gov (United States)

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Development of a transport cask for spent fuel elements of research reactors

International Nuclear Information System (INIS)

Quintana, F.; Saliba, R.O.; Furnari, J.C.; Mourao, R.P; Leite da Silva, L.; Novara, O.; Alexandre Miranda, C.; Mattar Neto, M.

2012-01-01

This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

Heterogeneous Recycle of Transuranics Fuels in Fast Reactors

International Nuclear Information System (INIS)

Hoffman, Edward; Taiwo, Temitope; Hill, Robert

2008-01-01

A preliminary physics evaluation of the impacts of heterogeneous recycle using Pu+Np driver and minor actinide target fuel assemblies in fast reactor cores has been performed by comparing results to those obtained for a reference homogeneous recycle core using driver assemblies containing grouped transuranic (TRU) fuel. Parametric studies are performed on the reference heterogeneous recycle core to evaluate the impacts of variations in the pre- and post-separation cooling times, target material type (uranium and non-uranium based), target amount and location, and other parameters on the system performance. This study focused on startup, single-pass cores for the purpose of quantifying impacts and also included comparisons to the option of simply storing the LWR spent nuclear fuel over a 50-year period. An evaluation of homogeneous recycle cores with elevated minor actinide contents is presented to illustrate the impact of using progressively higher TRU content on the core and transmutation performance, as a means of starting with known fuel technology with the aim of ultimately employing grouped TRU fuel in such cores. Reactivity coefficients and safety parameters are presented to indicate that the cores evaluated appear workable from a safety perspective, though more detailed safety and systems evaluations are required. (authors)

Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

Science.gov (United States)

George, Jeffrey A.

1991-01-01

A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

Feasibility of a Modified E-PASS and POSSUM System for Postoperative Risk Assessment in Patients with Spinal Disease.

Science.gov (United States)

Chun, Dong Hyun; Kim, Do Young; Choi, Sun Kyu; Shin, Dong Ah; Ha, Yoon; Kim, Keung Nyun; Yoon, Do Heum; Yi, Seong

2018-04-01

This retrospective case control study aimed to evaluate the feasibility of using Estimation of Physiological Ability and Surgical Stress (E-PASS) and Physiological and Operative Severity Score for the enumeration of Mortality and Morbidity (POSSUM) systems in patients undergoing spinal surgical procedures. Degenerative spine disease has increased in incidence in aging societies, as has the number of older adult patients undergoing spinal surgery. Many older adults are at a high surgical risk because of comorbidity and poor general health. We retrospectively reviewed 217 patients who had undergone spinal surgery at a single tertiary care. We investigated complications within 1 month after surgery. Criteria for both skin incision in E-PASS and operation magnitude in the POSSUM system were modified to fit spine surgery. We calculated the E-PASS and POSSUM scores for enrolled patients, and investigated the relationship between postoperative complications and both surgical risk scoring systems. To reinforce the predictive ability of the E-PASS system, we adjusted equations and developed modified E-PASS systems. The overall complication rate for spinal surgery was 22.6%. Forty-nine patients experienced 58 postoperative complications. Nineteen major complications, including hematoma, deep infection, pleural effusion, progression of weakness, pulmonary edema, esophageal injury, myocardial infarction, pneumonia, reoperation, renal failure, sepsis, and death, occurred in 17 patients. The area under the receiver operating characteristic curve (AUC) for predicted postoperative complications after spine surgery was 0.588 for E-PASS and 0.721 for POSSUM. For predicted major postoperative complications, the AUC increased to 0.619 for E-PASS and 0.842 for POSSUM. The AUC of the E-PASS system increased from 0.588 to 0.694 with the Modified E-PASS equation. The POSSUM system may be more useful than the E-PASS system for estimating postoperative surgical risk in patients undergoing

A First-Order One-Pass CPS Transformation

DEFF Research Database (Denmark)

Danvy, Olivier; Nielsen, Lasse Reichstein

2002-01-01

We present a new transformation of call-by-value lambdaterms into continuation-passing style (CPS). This transformation operates in one pass and is both compositional and first-order. Because it operates in one pass, it directly yields compact CPS programs that are comparable to what one would...... write by hand. Because it is compositional, it allows proofs by structural induction. Because it is first-order, reasoning about it does not require the use of a logical relation. This new CPS transformation connects two separate lines of research. It has already been used to state a new and simpler...... correctness proof of a direct-style transformation, and to develop a new and simpler CPS transformation of control-flow information....

Preliminary analysis of basic reactor physics of the Dual Fluid Reactor - 15270

International Nuclear Information System (INIS)

Wang, X.; Macian-Juan, R.; Seidl, M.

2015-01-01

The Dual Fluid Reactor (DFR) is a novel fast nuclear reactor concept invented by the IFK based on the Generation IV Molten Salt Reactor and the Liquid Metal Cooled Reactor. The DFR uses a chloride based molten fuel salt in order to harden the neutron spectrum. The molten fuel salt is cooled with a separated liquid lead loop, which in principle allows for higher power densities and better breeding performance. The DFR does not combine heat removal and breeding into a single circuit but separates the two functions into two independent circuits. Since there are attractive features mentioned in this design, the main task of this paper is to verify the model of the whole reactor based on this concept. For this purpose several calculations are presented, including steady state calculations, sensitivity calculations with regard to the nuclide cross sections, the temperature and geometry coefficient of k eff as well as the burnup calculation. The Monte Carlo calculation codes MCNP, SERPENT and SCALE are used for the analysis. As expected the study shows a significant negative reactivity feedback with temperature in the overall fission zone. For the coupled coolant and reflector design the temperature feedback is rather small for practical purposes such as reactor control during normal operation. In the view of these results the DFR in principle can be self-regulated totally by the temperature change of its own fuel salt and consequently can rely on fully passive safety systems for accident management

Fast reactor sodium systems operation experience and 'leak-before-break' criterion

International Nuclear Information System (INIS)

Ivanenko, V.N.; Zybin, V.A.

1996-01-01

In the paper sodium leakage detection systems used at fast reactors are described. Requirements on their main characteristics (sensitivity, response lime) are formulated. Results of tests are presented on studying the parameters of sodium leak detection systems including experiments on the measurement of size distribution of aerosol particles that have passed through sodium systems thermal insulation after leak initiation. Comparison of these data with dispersion of particles formed at free burning is carried out. Experience of real leaks that occurred at fast reactor sodium systems is analyzed. It has been shown that initiation and development of real leaks do not always follow the theoretical scheme. A substantial role of human factor for sodium systems reliability relative to sodium leaks is stressed. (author)

Reactor scale modeling of multi-walled carbon nanotube growth

International Nuclear Information System (INIS)

Lombardo, Jeffrey J.; Chiu, Wilson K.S.

2011-01-01

As the mechanisms of carbon nanotube (CNT) growth becomes known, it becomes important to understand how to implement this knowledge into reactor scale models to optimize CNT growth. In past work, we have reported fundamental mechanisms and competing deposition regimes that dictate single wall carbon nanotube growth. In this study, we will further explore the growth of carbon nanotubes with multiple walls. A tube flow chemical vapor deposition reactor is simulated using the commercial software package COMSOL, and considered the growth of single- and multi-walled carbon nanotubes. It was found that the limiting reaction processes for multi-walled carbon nanotubes change at different temperatures than the single walled carbon nanotubes and it was shown that the reactions directly governing CNT growth are a limiting process over certain parameters. This work shows that the optimum conditions for CNT growth are dependent on temperature, chemical concentration, and the number of nanotube walls. Optimal reactor conditions have been identified as defined by (1) a critical inlet methane concentration that results in hydrogen abstraction limited versus hydrocarbon adsorption limited reaction kinetic regime, and (2) activation energy of reaction for a given reactor temperature and inlet methane concentration. Successful optimization of a CNT growth processes requires taking all of those variables into account.

Nuclear reactor pressure vessel with an inner metal coating covered with a high temperature resistant thermal insulator

International Nuclear Information System (INIS)

1974-01-01

The thermal insulator covering the metal coating of a reactor vessel is designed for resisting high temperatures. It comprises one or several porous layers of ceramic fibers or of stacked metal foils, covered with a layer of bricks or ceramic tiles. The latter are fixed in position by fasteners comprising pins fixed to the coating and passing through said porous layers and fasteners (nut or bolts) for individually fixing the bricks to said pins, whereas ceramic plugs mounted on said bricks or tiles provide for the thermal insulation of the pins and of the nuts or bolts; such a thermal insulation can be applied to high-temperature reactors or to fast reactors [fr

Space reactors - What is a kilogram

International Nuclear Information System (INIS)

Buden, D.; Angelo, J.; Ek, D.; Voss, S.

1984-01-01

The use of nuclear electric propulsion can triple the payloads to GEO for a single Shuttle launch. Life orbits of 300 years can be used to allow most of the fission and activation products to decay before a reactor reenters the biosphere. Enough radioactive materials remain with very long lifetimes to make it desirable to design the reactor to disperse upon reentry and little additional risk to the biosphere is introduced by initiating NEP operations from 300 km

Space reactors: What is a kilogram

International Nuclear Information System (INIS)

Buden, D.; Angelo, J. Jr.; Ek, D.; Voss, S.

1984-01-01

The use of nuclear electric propulsion can triple payloads to GEO for a single Shuttle launch. Life orbits of 300 years can be used to allow most of the fission and activation products to decay before a reactor reenters the biosphere. Enough radioactive materials remain with very long lifetimes to make it desirable to design the reactor to disperse upon reentry and little additional risk to the biosphere is introduced by initiating NEP operations from 300 km

History, Development and Future of TRIGA Research Reactors

International Nuclear Information System (INIS)

2016-01-01

Due to its particular fuel design and resulting enhanced inherent safety features, TRIGA reactors (Training, Research, Isotopes, General Atomics) constitute a ‘class of their own’ among the large variety of research reactors built world-wide. This publication summarizes in a single document the information on the past and present of TRIGA research reactors and presents an outlook in view of potential issues to be solved by TRIGA operating organizations in the near future. It covers the historical development and basic TRIGA characteristics, followed by utilization, fuel conversion and ageing management of TRIGA research reactors. It continues with issues and challenges, introduction to the global TRIGA research reactor network and concludes with future perspectives. The publication is complemented with a CD-ROM to illustrate the historical developments of TRIGA research reactors through individual facility examples and experiences

Operational trials of single- and multi-element CR-39 dosemeters for the DIDO and PLUTO reactors at the Harwell Laboratory

International Nuclear Information System (INIS)

Gallacher, G.G.; Perks, C.A.

1993-01-01

Single- and multi-element CR-39 dosemeters, developed at the Harwell Laboratory, and a commercially available multi-element CR-39 dosemeter (obtained from Track Analysis Systems Ltd), were evaluated for their potential as neutron dosemeters for personnel working at Harwell Laboratory's research reactors. Owing to the angular dependence of the CR-39 (processed using electrochemical etching), the single-element dosemeter was found to be impractical. Consequently, a multi-element dosemeter was developed, which consisted of a cube of side 36 mm with CR-39 elements (also processed using electrochemical etching) attached to each of the sides. Although this dosemeter was technically suitable for this type of dosimetry, it was considered to be unacceptably bulky in personnel trials. The commercially available CR-39 dosemeter tested was much smaller (the CR-39 was only chemically etched) and this was considered to be acceptable as a personnel dosemeter. In addition, trials with personnel working at active handling glove boxes indicated that single-element dosemeters might be adequate, but further work would be needed to verify this. (author)

Coaxial higher-order mode damper employing a high-pass filter

International Nuclear Information System (INIS)

Kang, Y.W.; Jiang, X.

1997-01-01

Two different types of coaxial higher-order mode (HOM) dampers have been investigated for the Advanced Photon Source (APS) storage ring cavities: e-probe dampers and h-loop dampers. Realization of the h-loop dampers has been difficult because the loop antenna couples not only to the HOMs but also to the accelerating mode and results in loss of Q at the fundamental frequency. Previously, a first-order fundamental rejection filter was tested with unsatisfactory rejection characteristics. This problem can be overcome by using a higher-order high-pass filter between the loop and the matched load. Prototype dampers have been fabricated and tested in a storage ring single-cell cavity and the damping characteristic was analyzed

Survival and Passage of Juvenile Chinook Salmon and Steelhead Passing through Bonneville Dam, 2010

Energy Technology Data Exchange (ETDEWEB)

Ploskey, Gene R.; Weiland, Mark A.; Hughes, James S.; Woodley, Christa M.; Deng, Zhiqun; Carlson, Thomas J.; Kim, Jin A.; Royer, Ida M.; Batten, George W.; Cushing, Aaron W.; Carpenter, Scott M.; Etherington, D. J.; Faber, Derrek M.; Fischer, Eric S.; Fu, Tao; Hennen, Matthew J.; Mitchell, Tyler; Monter, Tyrell J.; Skalski, John R.; Townsend, Richard L.; Zimmerman, Shon A.

2011-12-01

Pacific Northwest National Laboratory (PNNL) and subcontractors conducted an acoustic-telemetry study of juvenile salmonid fish passage and survival at Bonneville Dam in 2010. The study was conducted to assess the readiness of the monitoring system for official compliance studies under the 2008 Biological Opinion and Fish Accords and to assess performance measures including route-specific fish passage proportions, travel times, and survival based upon a single-release model. This also was the last year of evaluation of effects of a behavioral guidance device installed in the Powerhouse 2 forebay. The study relied on releases of live Juvenile Salmon Acoustic Telemetry System tagged smolts in the Columbia River and used acoustic telemetry to evaluate the approach, passage, and survival of passing juvenile salmon. This study supports the U.S. Army Corps of Engineers continual effort to improve conditions for juvenile anadromous fish passing through Columbia River dams.

Survival and Passage of Juvenile Chinook Salmon and Steelhead Passing Through Bonneville Dam, 2010

Energy Technology Data Exchange (ETDEWEB)

Ploskey, Gene R.; Weiland, Mark A.; Hughes, James S.; Woodley, Christa M.; Deng, Zhiqun; Carlson, Thomas J.; Kim, Jin A.; Royer, Ida M.; Batten, George W.; Cushing, Aaron W.; Carpenter, Scott M.; Etherington, D. J.; Faber, Derrek M.; Fischer, Eric S.; Fu, Tao; Hennen, Matthew J.; Mitchell, T. D.; Monter, Tyrell J.; Skalski, J. R.; Townsend, Richard L.; Zimmerman, Shon A.

2012-09-01

Pacific Northwest National Laboratory (PNNL) and subcontractors conducted an acoustic-telemetry study of juvenile salmonid fish passage and survival at Bonneville Dam in 2010. The study was conducted to assess the readiness of the monitoring system for official compliance studies under the 2008 Biological Opinion and Fish Accords and to assess performance measures including route-specific fish passage proportions, travel times, and survival based upon a single-release model. This also was the last year of evaluation of effects of a behavioral guidance device installed in the Powerhouse 2 forebay. The study relied on releases of live Juvenile Salmon Acoustic Telemetry System tagged smolts in the Columbia River and used acoustic telemetry to evaluate the approach, passage, and survival of passing juvenile salmon. This study supports the U.S. Army Corps of Engineers continual effort to improve conditions for juvenile anadromous fish passing through Columbia River dams.

HETERO code, heterogeneous procedure for reactor calculation

International Nuclear Information System (INIS)

Jovanovic, S.M.; Raisic, N.M.

1966-11-01

This report describes the procedure for calculating the parameters of heterogeneous reactor system taking into account the interaction between fuel elements related to established geometry. First part contains the analysis of single fuel element in a diffusion medium, and criticality condition of the reactor system described by superposition of elements interactions. the possibility of performing such analysis by determination of heterogeneous system lattice is described in the second part. Computer code HETERO with the code KETAP (calculation of criticality factor η n and flux distribution) is part of this report together with the example of RB reactor square lattice

Integrated Cu-based TM-pass polarizer using CMOS technology platform

KAUST Repository

Ng, Tien Khee; Khan, Mohammed Zahed Mustafa; Ooi, Boon S.

2010-01-01

A transverse-magnetic-pass (TM-pass) copper (Cu) polarizer is proposed and analyzed using the previously published two-dimensional Method-of-Lines beam-propagation model. The proposed polarizer exhibits a simulated high-pass filter characteristics

Double-pass quantum volume hologram

International Nuclear Information System (INIS)

Vasilyev, Denis V.; Sokolov, Ivan V.

2011-01-01

We propose a scheme for parallel, spatially multimode quantum memory for light. The scheme is based on the propagation in different directions of a quantum signal wave and strong classical reference wave, like in a classical volume hologram and the previously proposed quantum volume hologram [D. V. Vasilyev et al., Phys. Rev. A 81, 020302(R) (2010)]. The medium for the hologram consists of a spatially extended ensemble of cold spin-polarized atoms. In the absence of the collective spin rotation during the interaction, two passes of light for both storage and retrieval are required, and therefore the present scheme can be called a double-pass quantum volume hologram. The scheme is less sensitive to diffraction and therefore is capable of achieving a higher density of storage of spatial modes as compared to the previously proposed thin quantum hologram [D. V. Vasilyev et al., Phys. Rev. A 77, 020302(R) (2008)], which also requires two passes of light for both storage and retrieval. However, the present scheme allows one to achieve a good memory performance with a lower optical depth of the atomic sample as compared to the quantum volume hologram. A quantum hologram capable of storing entangled images can become an important ingredient in quantum information processing and quantum imaging.

THE MODELING OF COUNTER-ROTATING TWIN-SCREW EXTRUDERS AS REACTORS FOR SINGLE-COMPONENT REACTIONS

NARCIS (Netherlands)

GANZEVELD, KJ; CAPEL, JE; VANDERWAL, DJ; JANSSEN, LPBM

Numerical models are useful to study the behaviour of the extruder as a polymerization reactor. With a correct numerical model a theoretical analysis of the influence of several reaction and extruder parameters can be made, the limitations of the use of the extruder reactor can be determined and the

Application programming interface document for the modernized Transient Reactor Analysis Code (TRAC-M)

International Nuclear Information System (INIS)

Mahaffy, J.; Boyack, B.E.; Steinke, R.G.

1998-05-01

The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization

Modeling drivers' passing duration and distance in a virtual environment

Directory of Open Access Journals (Sweden)

Haneen Farah

2013-07-01

The main contribution of this paper is in the empirical models developed for passing duration and distance which highlights the factors that affect drivers' passing behavior and can be used to enhance the passing models in simulation programs.

Dynamic model for tritium build-up at NPP with RBMK type reactors and its enviromental beraviour

International Nuclear Information System (INIS)

Badyaev, V.V.; Egorov, Yu.A.; Ivanov, E.A.; Stegachev, G.F.; Tolstykh, V.D.

1982-01-01

A model of tritium production dynamics for a high power NPP with RBMK type reactors is proposed and investigated. The main ''skeleton'' model structure for forecasting tritium buildup at a NPP and its exchange with the environment has been singled out at a heuristic level. Decomposition and layout of the units have been performed by global functional relations of the investigated objects (NPP and environment). the model accounts for only oxidized tritium forms. Water exchange between the NPP subsystems and environment is the main mechanism for tritium migration. The model does not account for scheduled periodic maintenance work effects, presence of stagnant zones in the station circuits, fuel burn-up, etc. The parametric identification method applied in the model makes the model adaptable to particular situations and considered systems of the NPP and environment. Completing the model with necessary and sufficient experimental data one can pass to certain forecasting problems and to NPP control as a tritium source in the environment

Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

Energy Technology Data Exchange (ETDEWEB)

Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

1997-11-01

A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

On the influence of high-pass filtering on ICA-based artifact reduction in EEG-ERP.

Science.gov (United States)

Winkler, Irene; Debener, Stefan; Müller, Klaus-Robert; Tangermann, Michael

2015-01-01

Standard artifact removal methods for electroencephalographic (EEG) signals are either based on Independent Component Analysis (ICA) or they regress out ocular activity measured at electrooculogram (EOG) channels. Successful ICA-based artifact reduction relies on suitable pre-processing. Here we systematically evaluate the effects of high-pass filtering at different frequencies. Offline analyses were based on event-related potential data from 21 participants performing a standard auditory oddball task and an automatic artifactual component classifier method (MARA). As a pre-processing step for ICA, high-pass filtering between 1-2 Hz consistently produced good results in terms of signal-to-noise ratio (SNR), single-trial classification accuracy and the percentage of `near-dipolar' ICA components. Relative to no artifact reduction, ICA-based artifact removal significantly improved SNR and classification accuracy. This was not the case for a regression-based approach to remove EOG artifacts.

Field Performance Test of an Air-Cleaner with Photocatalysis-Plasma Synergistic Reactors for Practical and Long-Term Use

Directory of Open Access Journals (Sweden)

Tsuyoshi Ochiai

2014-10-01

Full Text Available A practical and long-term usable air-cleaner based on the synergy of photocatalysis and plasma treatments has been developed. A field test of the air-cleaner was carried out in an office smoking room. The results were compared to previously reported laboratory test results. Even after a treatment of 12,000 cigarettes-worth of tobacco smoke, the air-cleaner maintained high-level air-purification activity (98.9% ± 0.1% and 88% ± 1% removal of the total suspended particulate (TSP and total volatile organic compound (TVOC concentrations, respectively at single-pass conditions. Although the removal ratio of TSP concentrations was 98.6% ± 0.2%, the ratio of TVOC concentrations was 43.8% after a treatment of 21,900 cigarettes-worth of tobacco smoke in the field test. These results indicate the importance of suitable maintenance of the reactors in the air-cleaner during field use.

Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment

International Nuclear Information System (INIS)

Zhu, Xiuping; Logan, Bruce E.

2013-01-01

Highlights: ► A new type of electro-Fenton system was developed for wastewater treatment. ► Degradation efficiency of organic pollutants was substantially improved. ► Operation cost was greatly reduced compared to other microbial fuel cell designs. -- Abstract: Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H 2 O 2 generation and Fe 2+ release for the Fenton reaction. In tests using phenol, 75 ± 2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22 h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

Energy Technology Data Exchange (ETDEWEB)

1993-05-13

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

The effect of a single tensile overload on stress corrosion cracking growth of stainless steel in a light water reactor environment

International Nuclear Information System (INIS)

Xue He; Li Zhijun; Lu Zhanpeng; Shoji, Tetsuo

2011-01-01

Research highlights: → The affect of a single tensile overload on SCC growth rate is investigated. → A single tensile overload would produce a residual plastic strain in the SCC tip. → The residual plastic strain would decrease the plastic strain rate in the SCC tip. → A single tensile overload would cause crack growth rate retardation in the SCC tip. → SCC growth rate in the quasi-stationary crack tip is relatively lower. - Abstract: It has been found that a single tensile overload applied during constant load amplitude might cause crack growth rate retardation in various crack propagating experiments which include fatigue test and stress corrosion cracking (SCC) test. To understand the affecting mechanism of a single tensile overload on SCC growth rate of stainless steel or nickel base alloy in light water reactor environment, based on elastic-plastic finite element method (EPFEM), the residual plastic strain in both tips of stationary and growing crack of contoured double cantilever beam (CDCB) specimen was simulated and analyzed in this study. The results of this investigation demonstrate that a residual plastic strain in the region immediately ahead of the crack tips will be produced when a single tensile overload is applied, and the residual plastic strain will decrease the plastic strain rate level in the growing crack tip, which will causes crack growth rate retardation in the tip of SCC.

Experimental analysis of humidification process by air passing through seawater

International Nuclear Information System (INIS)

El-Agouz, S.A.; Abugderah, M.

2008-01-01

An experimental investigation of humidification process by air passing through seawater is presented. The main objective of this work was to determine the humid air behaviour through single-stage of heating-humidifying processes. This experimental work studied the influence of the operating conditions such as the water temperature, the headwater difference, the air velocity and the inlet air temperature to evaporator chamber on the vapour content difference and humidification efficiency. Two cases of different inlet conditions of ambient and heated air cases are studied. The experimental results show that, the vapour content difference and the humidification efficiency of the system is strongly affected by the saline water temperature in the evaporator chamber, headwater difference and the air velocity. The inlet air temperature to evaporator chamber variation was found to have a small affect on the vapour content difference. The obtained maximum vapour content difference of the air was about 222 gr w /kg a at 75 deg. C for water and air. The obtained vapour content is high compared to that obtained in literature for single-stage and very similar for multi-stage

Estimating the Exchange Rate Pass-Through to Prices in Mexico

OpenAIRE

Josué Fernando Cortés Espada

2013-01-01

This paper estimates the magnitude of the exchange rate pass-through to consumer prices in Mexico. Moreover, it analyzes if the pass-through dynamics have changed in recent years. In particular, it uses a methodology that generates results consistent with the hierarchy implicit in the cpi. The results suggest that the exchange rate pass-through to the general price level is low and not statistically significant. However, the pass-through is positive and significant for goods prices. Furthermo...

Analysis of Radioactivity Contamination Level of Kartini Reactor Efluen Gas to the Environment

International Nuclear Information System (INIS)

Suratman; Purwanto; Aminjoyo, S

1996-01-01

The analysis of radioactivity contamination level of Kartini reactor efluen gas to the environment has been done from 13-10-'95 until 8-2-'96. The aim of this research is to determine the radioactivity contamination level on the environment resulted from the release of Kartini reactor efluen gas and other facilities at Yogyakarta Nuclear Research Centre through stack. The analysis methods is the student t-test, the first count factor test and the gamma spectrometry. The gas sampling were carried out in the stack reactor, reactor room, environment and in other room for comparison. Efluen gas was sucked through a filter by a high volume vacuum pump. The filter was counted for beta, gamma and alpha activities. The radioactivity contamination level of the efluen gas passing through the stack to the environment was measured between 0.57 - 1.34 Bq/m3, which was equal to the airborne radioactivity in environment between 0.69 - 1.12 Bq/m3. This radioactivity comes from radon daughter, decay products result from the natural uranium and thorium series of the materials of the building

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

International Nuclear Information System (INIS)

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

2004-01-01

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