Characterization of Subsurface Defects in Ceramic Rods by Laser Scattering and Fractography

International Nuclear Information System (INIS)

Zhang, J. M.; Sun, J. G.; Andrews, M. J.; Ramesh, A.; Tretheway, J. S.; Longanbach, D. M.

2006-01-01

Silicon nitride ceramics are leading materials being evaluated for valve train components in diesel engine applications. The surface and subsurface defects and damage induced by surface machining can significantly affect component strength and lifetime. In this study, a nondestructive evaluation (NDE) technique based upon laser scattering has been utilized to analyze eight transversely ground silicon nitride cylindrical rods before fracture tests. The fracture origins (machining cracks or material-inherent flaws) identified by fractography after fracture testing were correlated with laser scattering images. The results indicate that laser scattering is able to identify possible fracture origin in the silicon nitride subsurface without the need for destructive fracture tests

Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

International Nuclear Information System (INIS)

Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

2013-01-01

It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

Laser amplifier based on a neodymium glass rod 150 mm in diameter

Energy Technology Data Exchange (ETDEWEB)

Shaykin, A A; Fokin, A P; Soloviev, A A; Kuzmin, A A; Shaikin, I A; Burdonov, K F; Khazanov, E A [Institute of Applied Physics, Russian Academy of Sciences, Nizhnii Novgorod (Russian Federation); Charukhchev, A V [Public Limited Company " Scientific research Institute for Optoelectronic Instrument Engineering" , Leningrad region (Russian Federation)

2014-05-30

A unique large-aperture neodymium glass rod amplifier is experimentally studied. The small-signal gain distribution is measured at different pump energies. The aperture-averaged gain is found to be 2.3. The stored energy (500 J), the maximum possible pump pulse repetition rate, and the depolarisation in a single pulse and in a series of pulses with a repetition rate of one pulse per five minutes are calculated based on the investigations performed. It is shown that the use of this amplifier at the exit of the existing laser can increase the output pulse energy from 300 to 600 J. (lasers)

Simple measuring rod method for the coaxiality of serial holes

Science.gov (United States)

Wang, Lei; Yang, Tongyu; Wang, Zhong; Ji, Yuchen; Liu, Changjie; Fu, Luhua

2017-11-01

Aiming at the rapid coaxiality measurement of serial hole part with a small diameter, a coaxiality measuring rod for each layer hole with a single LDS (laser displacement sensor) is proposed. This method does not require the rotation angle information of the rod, and the coaxiality of serial holes can be calculated from the measured values of LDSs after randomly rotating the measuring rod several times. With the mathematical model of the coaxiality measuring rod, each factor affecting the accuracy of coaxiality measurement is analyzed by simulation, and the installation accuracy requirements of the measuring rod and LDSs are presented. In the tolerance of a certain installation error of the measuring rod, the relative center of the hole is calculated by setting the over-determined nonlinear equations of the fitting circles of the multi-layer holes. In experiment, coaxiality measurement accuracy is realized by a 16 μm precision LDS, and the validity of the measurement method is verified. The manufacture and measurement requirements of the coaxiality measuring rod are low, by changing the position of LDSs in the measuring rod, the serial holes with different sizes and numbers can be measured. The rapid coaxiality measurement of parts can be easily implemented in industrial sites.

Control rod cluster with removable rods for nuclear fuel assembly

International Nuclear Information System (INIS)

Denizou, J.P.

1989-01-01

For each removable control rod, the open end section of the sleeve has a certain length of reduced diameter with openings in its wall. The top end of the rod is joined to an extension tube that surrounds the shaft over part of its lenght. This extension tube fits over the reduced part of the sleeve when the shaft is screwed into the bore of the sleeve. Rotation of the rod in the sleeve is prevented by deforming the extension tube locally in the openings of the end part of the sleeve. The rod is dismantled by exerting a torque on it using a gripping area near the end of the rod [fr

Experimental data base of turbulent flow in rod bundles using laser doppler velocimeter

International Nuclear Information System (INIS)

Chung, Moon Ki; Yang, Sun Kyu; Chung, Heung June; Won, Soon Yeun; Kim, Bok Deuk; Cho, Young Rho

1992-01-01

This report presents in detail the hydraulic characteristics measurements in subchannels of rod bundles using one-component LDV (Laser Doppler Velocimeter). In particular, this report presents the figures and tabulations of the resulting data. The detailed explanations about these results are shown in references publicated or presented at the conference. 4 kinds of experimental work were performed so far. (Author)

Rod-like microglia are restricted to eyes with laser-induced ocular hypertension but absent from the microglial changes in the contralateral untreated eye.

Directory of Open Access Journals (Sweden)

Rosa de Hoz

Full Text Available In the mouse model of unilateral laser-induced ocular hypertension (OHT the microglia in both the treated and the normotensive untreated contralateral eye have morphological signs of activation and up-regulation of MHC-II expression in comparison with naïve. In the brain, rod-like microglia align to less-injured neurons in an effort to limit damage. We investigate whether: i microglial activation is secondary to laser injury or to a higher IOP and; ii the presence of rod-like microglia is related to OHT. Three groups of mice were used: age-matched control (naïve, n=15; and two lasered: limbal (OHT, n=15; and non-draining portion of the sclera (scleral, n=3. In the lasered animals, treated eyes as well as contralateral eyes were analysed. Retinal whole-mounts were immunostained with antibodies against, Iba-1, NF-200, MHC-II, CD86, CD68 and Ym1. In the scleral group (normal ocular pressure no microglial signs of activation were found. Similarly to naïve eyes, OHT-eyes and their contralateral eyes had ramified microglia in the nerve-fibre layer related to the blood vessel. However, only eyes with OHT had rod-like microglia that aligned end-to-end, coupling to form trains of multiple cells running parallel to axons in the retinal surface. Rod-like microglia were CD68+ and were related to retinal ganglion cells (RGCs showing signs of degeneration (NF-200+RGCs. Although MHC-II expression was up-regulated in the microglia of the NFL both in OHT-eyes and their contralateral eyes, no expression of CD86 and Ym1 was detected in ramified or in rod-like microglia. After 15 days of unilateral lasering of the limbal and the non-draining portion of the sclera, activated microglia was restricted to OHT-eyes and their contralateral eyes. However, rod-like microglia were restricted to eyes with OHT and degenerated NF-200+RGCs and were absent from their contralateral eyes. Thus, rod-like microglia seem be related to the neurodegeneration associated with HTO.

Lasing characteristics of refractive-index-matched composite Y3Al5O12 rods employing transparent ceramics for solar-pumped lasers

Science.gov (United States)

Hasegawa, Kazuo; Ichikawa, Tadashi; Takeda, Yasuhiko; Ikesue, Akio; Ito, Hiroshi; Motohiro, Tomoyoshi

2018-04-01

We have proposed a new configuration of solar-pumped lasers employing transparent ceramic rods. The laser rod has a composite structure consisting of a Nd/Cr:YAG gain domain surrounded by Gd:YAG with the same refractive index as that of Nd/Cr:YAG. The lasing mode is well controlled by the output coupler, and the parasitic oscillation is suppressed, owing to the refractive index matching. A high laser slope efficiency and a low laser oscillation threshold were achieved owing to the suppressed absorption outside the lasing mode, which was previously a serious issue for the end-pumping configuration using a high-NA focusing optics. The laser oscillation threshold of 136 mW and the slope efficiency of 25.3% were derived. Thus, we have resolved the issue of useless absorption associated with the high-NA end-pumping, and achieved significant improvements compared with the conventional structure of uniform Nd/Cr:YAG.

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

International Nuclear Information System (INIS)

Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

2014-01-01

A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

Photonic crystal rod fibers: Understanding a new class of active optical waveguides

DEFF Research Database (Denmark)

Laurila, Marko

core modes and the cladding band as the underlying mechanics to ensure SM operation of the new rod fiber design (85 μm core diameter), which was developed during this thesis work. The second focus of this work is the study of the new ytterbiumdoped rod fiber design under active operation. Performance...... of the rod fiber is evaluated in high power laser and laser amplifier configurations. The high power rod amplifier setup including the seed source is developed and characterized. Results obtained from the rod fiber showed simultaneously SM, near diffraction limited output beam quality with high average power...... and pulse energy generation using both laser and laser amplifier configurations. Modal instabilities (MIs) in high power fiber amplifiers are discussed, and a memory effect of the MI threshold level together with a recovery method and evidence of improved performance while suppressing MIs are reported...

Internal attachment of laser beam welded stainless steel sheathed thermocouples into stainless steel upper end caps in nuclear fuel rods for the LOFT Reactor

International Nuclear Information System (INIS)

Welty, R.K.; Reid, R.D.

1980-01-01

The Exxon Nuclear Company, Inc., acting as a subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, conducted a laser beam welding study to attach internal stainless steel thermocouples into stainless steel upper end caps in nuclear fuel rods. The objective of this study was to determine the feasibility of laser welding a single 0.063 inch diameter stainless steel (304) sheathed thermocouple into a stainless steel (316) upper end cap for nuclear fuel rods. A laser beam was selected because of the extremely high energy input in unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A special weld fixture was designed and fabricated to hold the end cap and the thermocouple with angular and rotational adjustment under the laser beam. A commercial pulsed laser and energy control system was used to make the welds

Hybrid fiber-rod laser

Science.gov (United States)

Beach, Raymond J.; Dawson, Jay W.; Messerly, Michael J.; Barty, Christopher P. J.

2012-12-18

Single, or near single transverse mode waveguide definition is produced using a single homogeneous medium to transport both the pump excitation light and generated laser light. By properly configuring the pump deposition and resulting thermal power generation in the waveguide device, a thermal focusing power is established that supports perturbation-stable guided wave propagation of an appropriately configured single or near single transverse mode laser beam and/or laser pulse.

Control rod assembly

International Nuclear Information System (INIS)

Takahashi, Akio.

1982-01-01

Purpose: To enable reliable insertion and drops of control rods, as well as insure a sufficient flow rate of coolants flowing through the control rods for attaining satisfactory cooling thereof to enable relexation of thermal stress resulted to rectifying mechanisms or the likes. Constitution: To the outer circumference of a control rod contained vertically movably within a control rod guide tube, resistive members are retractably provided in such a way as to project to close the gap between outer circumference of the control rod and the inner surface of the control rod guide tube upon engagement of a gripper of control rod drives, and retract upon release of the engagement of the gripper. Thus, since the resistive members project to provide a greater resistance to the coolants flowing between them and the control rod guide tube in the normal operation where the gripper is engaged to drive the control rod by the control rod drives, a major part of the coolant flowing into the control rod guide tube flows into the control rod. This enables to cool the control rod effectively and make the temperature distribution uniform for the coolant flowing from the upper end of the control rod guide tube to thereby attain the relaxation of the thermal stress resulted in the rectifying mechanisms or the likes. (Moriyama, K.)

Process and equipment for pressure build-up in nuclear reactor fuel rods

International Nuclear Information System (INIS)

Heer, W.F.; Carli, E.V. de.

1976-01-01

The equipment makes possible the build-up of inert gas pressure in a filled and closed fuel can, i.e. in a complete fuel rod. Handling is simple, it is suitable for mass production and only causes low processing costs. The quality, e.g. the degree of purity of the contents of the rod, remains unchangedin processing. The equipment consists of a vacuum-tight space, into which the equally vacuum tight fuel rod is introduced, and can be fixed so that its position can be reproduced unmistakeably. The vacuum space contains a connection for the inert gases and a laser arrangement. After inserting a fuel rod into the facility, this is evacuated and the fuel can has a hole bored in it by a laser beam. After fast equalisation of pressure, an inert gas at the required pressure is introduced into the chamber and the fuel rod. After the filling process is completed, the fuel can is closed again with the same laser beam. The quality of the seal obtained, i.e the leak-tightness of the fuel can, can be checked after reduction of the inert gas pressure and before taking out the fuel rod, by repeated evacuation of the chamber. Laser light energies between 13,000 and 110,000 Joule/sq cm are sufficient. Optimum results were obtained for a Zircaloy fuel can with about 52,000 Joule/sq cm. (TK) [de

Clinical results from low-level laser therapy in patients with autosomal dominant cone-rod dystrophy

Science.gov (United States)

Koev, K.; Avramov, L.; Borissova, E.

2018-03-01

The objective of this study is to examine long-term effects of low-level laser therapy (LLLT) in patients with autosomal dominant cone-rod dystrophy (CRDs). A He-Ne Laser with continuous emission at 633 nm (01 mW/cm2) was used on five patients with autosomal dominant pedigree of Romani origin with non-syndromic CRDs. The laser radiation was applied transpupillary to the macula six times for three minutes every other day. The experiment was conducted for a period of three years. The clinical evaluation included best corrected visual acuity determination, funduscopy, Humphrey perimetry, Farnsworth Hue-28 color testing, fluorescein angiography, and full-field electroretinogram (ERG). All affected individuals presented reduced visual acuity (0.01 – 0.4) and photophobia. The funduscopic examination and fluorescein angiography revealed advanced changes including bone spicule-like pigment deposits in the midperiphery and the macular area, along with retinal atrophy, narrowing of the vessels, and waxy optic discs. The visual fields demonstrated central scotoma. The electrophysiologic examination of the patients detected an abnormal cone-rod ERG (20 – 30 μV) with photopic amplitudes more markedly reduced than the scotopic. Flicker responses were missing and Farnsworth Hue-28 test found protanopia. There was a statistically significant increase in the visual acuity (p<0.001, end of study versus baseline) for CRDs patients for the period of three years after the treatment with LLLT. Following the LLLT, the central absolute scotoma in CRDs was reduced, as was the prevalence of metamorphopsia in CRDs. This study shows that LLLT may prove be a novel long-lasting therapeutic option for both forms of CRDs. It is a highly effective treatment resulting in a long-term improvement of the visual acuity.

Variation in sensitivity, absorption and density of the central rod distribution with eccentricity.

Science.gov (United States)

Tornow, R P; Stilling, R

1998-01-01

To assess the human rod photopigment distribution and sensitivity with high spatial resolution within the central +/-15 degrees and to compare the results of pigment absorption, sensitivity and rod density distribution (number of rods per square degree). Rod photopigment density distribution was measured with imaging densitometry using a modified Rodenstock scanning laser ophthalmoscope. Dark-adapted sensitivity profiles were measured with green stimuli (17' arc diameter, 1 degrees spacing) using a T ubingen manual perimeter. Sensitivity profiles were plotted on a linear scale and rod photopigment optical density distribution profiles were converted to absorption profiles of the rod photopigment layer. Both the absorption profile of the rod photopigment and the linear sensitivity profile for green stimuli show a minimum at the foveal center and increase steeply with eccentricity. The variation with eccentricity corresponds to the rod density distribution. Rod photopigment absorption profiles, retinal sensitivity profiles, and the rod density distribution are linearly related within the central +/-15 degrees. This is in agreement with theoretical considerations. Both methods, imaging retinal densitometry using a scanning laser ophthalmoscope and dark-adapted perimetry with small green stimuli, are useful for assessing the central rod distribution and sensitivity. However, at present, both methods have limitations. Suggestions for improving the reliability of both methods are given.

Development of joining techniques for fabrication of fuel rod simulators

International Nuclear Information System (INIS)

Moorhead, A.J.; McCulloch, R.W.; Reed, R.W.; Woodhouse, J.J.

1980-10-01

Much of the safety-related thermal-hydraulic tests on nuclear reactors are conducted not in the reactor itself, but in mockup segments of a core that uses resistance-heated fuel rod simulators (FRS) in place of the radioactive fuel rods. Laser welding and furnace brazing techniques are described for joining subassemblies for FRS that have survived up to 1000 h steady-state operation at 700 to 1100 0 C cladding temperatures and over 5000 thermal transients, ranging from 10 to 100 0 C/s. A pulsed-laser welding procedure that includes use of small-diameter filler wire is used to join one end of a resistance heating element of Pt-8 W, Fe-22 Cr-5.5 Al-0.5 Co, or 80 Ni-20 Cr (wt %) to a tubular conductor of an appropriate intermediate material. The other end of the heating element is laser welded to an end plug, which in turn is welded to a central conductor rod

Hydrodynamic behavior of a bare rod bundle

International Nuclear Information System (INIS)

Bartzis, J.G.; Todreas, N.E.

1977-06-01

The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers

Study of laser cladding nuclear valve parts

International Nuclear Information System (INIS)

Shi Shihong; Wang Xinlin; Huang Guodong

1998-12-01

The mechanism of laser cladding is discussed by using heat transfer model of laser cladding, heat conduction model of laser cladding and convective transfer mass model of laser melt-pool. Subsequently the laser cladding speed limit and the influence of laser cladding parameters on cladding layer structure is analyzed. A 5 kW with CO 2 transverse flow is used in the research for cladding treatment of sealing surface of stop valve parts of nuclear power stations. The laser cladding layer is found to be 3.0 mm thick. The cladding surface is smooth and has no such defects as crack, gas pore, etc. A series of comparisons with plasma spurt welding and arc bead welding has been performed. The results show that there are higher grain grade and hardness, lower dilution and better performances of resistance to abrasion, wear and of anti-erosion in the laser cladding layer. The new technology of laser cladding can obviously improve the quality of nuclear valve parts. Consequently it is possible to lengthen the service life of nuclear valve and to raise the safety and reliability of the production system

Control rod for a nuclear reactor

Science.gov (United States)

Roman, Walter G.; Sutton, Jr., Harry G.

1979-01-01

A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod.

Control rod for a nuclear reactor

International Nuclear Information System (INIS)

Roman, W.G.; Sutton, H.G. Jr.

1976-01-01

A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilent members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

Control rod for a nuclear reactor

International Nuclear Information System (INIS)

Roman, W.G.; Sutton, H.G. Jr.

1979-01-01

A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

Microstructural engineering applied to the controlled cooling of steel wire rod: Part I. Experimental design and heat transfer

Science.gov (United States)

Campbell, P. C.; Hawbolt, E. B.; Brimacombe, J. K.

1991-11-01

The goal of this study was to develop a mathematical model which incorporates heat flow, phase transformation kinetics, and property-structure-composition relationships to predict the mechanical properties of steel rod being control cooled under industrial conditions. Thus, the principles of microstructural engineering have been brought to bear on this interdisciplinary problem by combining computer modeling with laboratory measurements of heat flow, austenite decomposition kinetics, microstructure and mechanical properties, and industrial trials to determine heat transfer and obtain rod samples under known conditions. Owing to the length and diversity of the study, it is reported in three parts,[8191]the first of which is concerned with the heat flow measurements. A relatively simple and reliable technique, involving a preheated steel rod instrumented with a thermocouple secured at its centerline, has been devised to determine the cooling rate in different regions of the moving bed of rod loops on an operating Stelmor line. The measured thermal response of the rod has been analyzed by two transient conduction models (lumped and distributed parameter, respectively) to yield overall heat-transfer coefficients for radiation and convection. The adequacy of the technique has been checked by cooling instrumented rods under well-defined, air crossflow conditions in the laboratory and comparing measured heat-transfer coefficients to values predicted from well-established equations. The industrial thermal measurements have permitted the characterization of a coefficient to account for radiative interaction among adjacent rod loops near the edge and at the center of the bed.

Hydrodynamic behavior of a bare rod bundle. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Bartzis, J.G.; Todreas, N.E.

1977-06-01

The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers.

Remote helium leak test of the DUPIC fuel rod

International Nuclear Information System (INIS)

Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

1998-01-01

DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

Optimized Control Rods of the BR2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, Silva; Koonen, E.

2007-09-15

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Optimized Control Rods of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, E.

2007-01-01

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Single-crystal SrTiO3 fiber grown by laser heated pedestal growth method: influence of ceramic feed rod preparation in fiber quality

Directory of Open Access Journals (Sweden)

D. Reyes Ardila

1998-10-01

Full Text Available The rapidly spreading use of optical fiber as a transmission medium has created an interest in fiber-compatible optical devices and methods for growing them, such as the Laser Heated Pedestal Growth (LHPG. This paper reports on the influence of the ceramic feed rod treatment on fiber quality and optimization of ceramic pedestal processing that allows improvements to be made on the final quality in a simple manner. Using the LHPG technique, transparent crack-free colorless single crystal fibers of SrTiO3 (0.50 mm in diameter and 30-40 mm in length were grown directly from green-body feed rods, without using external oxygen atmosphere.

Prototypical spent nuclear nuclear fuel rod consolidation equipment, Phase 2: Final design report: Volume 2, Appendices: Part 1

International Nuclear Information System (INIS)

Ciez, A.P.

1987-01-01

The purpose of this specification is to establish functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation System. The Department of Energy-Idaho Operations Office (DOE-ID) is responsible for the implementation of the Prototypic Dry Rod Consolidation Demonstration Project. This program is to develop and demonstrate a fully qualified, licensable, cost-effective, dry spent fuel rod consolidation system by July 1989. The work is divided into four phases as follows: Phase I--Preliminary Design, Phase II--Final Design Option, Phase III--Fabrication and System Checkout Option, and Phase IV--Installation and Hot Demonstration Option. This specification is part of the Phase II effort. The objectives of this specification are to provide functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation equipment; establish specific tool and subsystem requirements such that the integrated and overall system requirements are satisfied; and establish positioning, envelope and size interface control requirements for each tool or subsystem such that the individual components will interface properly with the overall system design

Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

International Nuclear Information System (INIS)

Malang, S.

1975-11-01

An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

Control rod homogenization in heterogeneous sodium-cooled fast reactors

International Nuclear Information System (INIS)

Andersson, Mikael

2016-01-01

The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

Application of contact mechanics for fretting damage of fuel rod: part 1 influence functions and numerical method

International Nuclear Information System (INIS)

Kim, H. K.; Yoon, K. H.; Kang, H. S.; Song, G. N.

1998-01-01

For the analysis of the fretting problem of the fuel rods, present paper(Part I) shows the numerical method developed for evaluating the stresses on the contact surfaces between the fuel rods and the spacer grids. Theory of Contact Mechanics was incorporated. Contact area was regarded as a plane strain condition, so plane problem was taken into consideration. Normal stress profile on the contact surface was assumed to be Hertzian. As for the direction of the shear load, a closed load path, e.g. load increase in transverse increase in axial decrease in transverse decrease in axial increase in transverse increase in axial direction was considered for simulating the rod vibration in a reactor core. Partial slip problem was consulted. As for the numerical method, a triangular traction element was utilized and the corresponding influence functions were evaluated. Numerical program has been implemented for the present analysis, of which the validity was verified by comparing the Mindlin-Cattaneo solution

Fuel followed control rod installation at AFRRI

International Nuclear Information System (INIS)

Moore, Mark; Owens, Chris; Forsbacka, Matt

1992-01-01

Fuel Followed Control Rods (FFCRs) were installed at the Armed Forces Radiobiology Research Institute's 1 MW TRIGA Reactor. The procedures for obtaining, shipping, and installing the FFCRs is described. As part of the FFCR installation, the transient rod drive was relocated. Core performance due to the addition of the fuel followed control rods is discussed. (author)

Control rod driving mechanism

International Nuclear Information System (INIS)

Ooshima, Yoshio.

1983-01-01

Purpose: To perform reliable scram operation, even if abnormality should occur in a system instructing scram operation in FBR type reactors. Constitution: An aluminum alloy member to be melt at a predetermined temperature (about 600sup(o)C) is disposed to a connection part between a control rod and a driving mechanism, whereby the control rod is detached from the driving mechanism and gravitationally fallen to the reactor core. (Ikeda, J.)

120 watt continuous wave solar-pumped laser with a liquid light-guide lens and an Nd:YAG rod.

Science.gov (United States)

Dinh, T H; Ohkubo, T; Yabe, T; Kuboyama, H

2012-07-01

We propose a simple and efficient pumping approach for a high-power solar-pumped laser by using a liquid light-guide lens (LLGL) and a hybrid pumping cavity. A 2×2 m Fresnel lens is used as a primary concentrator to collect natural sunlight; 120 W cw laser power and a 4.3% total slope efficiency are achieved with a 6-mm diameter Nd:YAG rod within a 14-mm diameter LLGL. The corresponded collection efficiency is 30.0 W/m(2), which is 1.5 times larger than the previous record. This result is unexpectedly better than that of Cr:Nd:YAG ceramics. It is because the scattering coefficient of Cr:Nd:YAG ceramics is 0.004cm(1), which is 2 times larger than that of the Nd:YAG crystal, although both have similar saturation gains.

Feasibility demonstration of using wire electrical-discharge machining, abrasive flow honing, and laser spot welding to manufacture high-precision triangular-pitch Zircaloy-4 fuel-rod-support grids

International Nuclear Information System (INIS)

Horwood, W.A.

1982-05-01

Results are reported supporting the feasibility of manufacturing high precision machined triangular pitch Zircaloy-4 fuel rod support grids for application in water cooled nuclear power reactors. The manufacturing processes investigated included wire electrical discharge machining of the fuel rod and guide tube cells in Zircaloy plate stock to provide the grid body, multistep pickling of the machined grid to provide smooth and corrosion resistant surfaces, and laser welding of thin Zircaloy cover plates to both sides of the grid body to capture separate AM-350 stainless steel insert springs in the grid body. Results indicated that dimensional accuracy better than +- 0.001 and +- 0.002 inch could be obtained on cell shape and position respectively after wire EDM and surface pickling. Results on strength, corrosion resistance, and internal quality of laser spot welds are provided

Some aspects of precise laser machining - Part 1: Theory

Science.gov (United States)

Wyszynski, Dominik; Grabowski, Marcin; Lipiec, Piotr

2018-05-01

The paper describes the role of laser beam polarization and deflection on quality of laser beam machined parts made of difficult to cut materials (used for cutting tools). Application of efficient and precise cutting tool (laser beam) has significant impact on preparation and finishing operations of cutting tools for aviation part manufacturing. Understanding the phenomena occurring in the polarized light laser cutting gave possibility to design, build and test opto-mechanical instrumentation to control and maintain process parameters and conditions. The research was carried within INNOLOT program funded by Polish National Centre for Research and Development.

Fuel rod fixing system

International Nuclear Information System (INIS)

Christiansen, D.W.

1982-01-01

This is a reusable system for fixing a nuclear reactor fuel rod to a support. An interlock cap is fixed to the fuel rod and an interlock strip is fixed to the support. The interlock cap has two opposed fingers, which are shaped so that a base is formed with a body part. The interlock strip has an extension, which is shaped so that this is rigidly fixed to the body part of the base. The fingers of the interlock cap are elastic in bending. To fix it, the interlock cap is pushed longitudinally on to the interlock strip, which causes the extension to bend the fingers open in order to engage with the body part of the base. To remove it, the procedure is reversed. (orig.) [de

Fuel rod puncturing and fission gas monitoring system examination techniques

International Nuclear Information System (INIS)

Song, Woong Sup

1999-02-01

Fission gas products accumulated in irradiated fuel rod is 1-2 cm 3 in CANDU and 40-50 cm 3 in PWR fuel rod. Fuel rod puncturing and fission gas monitoring system can be used for both CANDU and PWR fuel rod. This system comprises puncturing device located at in cell part and monitoring device located at out cell part. The system has computerized 9 modes and can calculate both void volume and mass volume only single puncturing. This report describes techniques and procedure for operating fuel rod puncturing and gas monitoring system which can be play an important role in successful operation of the devices. Results obtained from the analysis can give more influence over design for fuel rods. (Author). 6 refs., 9 figs

High-average-power diode-pumped Yb: YAG lasers

International Nuclear Information System (INIS)

Avizonis, P V; Beach, R; Bibeau, C M; Emanuel, M A; Harris, D G; Honea, E C; Monroe, R S; Payne, S A; Skidmore, J A; Sutton, S B

1999-01-01

A scaleable diode end-pumping technology for high-average-power slab and rod lasers has been under development for the past several years at Lawrence Livermore National Laboratory (LLNL). This technology has particular application to high average power Yb:YAG lasers that utilize a rod configured gain element. Previously, this rod configured approach has achieved average output powers in a single 5 cm long by 2 mm diameter Yb:YAG rod of 430 W cw and 280 W q-switched. High beam quality (M(sup 2)= 2.4) q-switched operation has also been demonstrated at over 180 W of average output power. More recently, using a dual rod configuration consisting of two, 5 cm long by 2 mm diameter laser rods with birefringence compensation, we have achieved 1080 W of cw output with an M(sup 2) value of 13.5 at an optical-to-optical conversion efficiency of 27.5%. With the same dual rod laser operated in a q-switched mode, we have also demonstrated 532 W of average power with an M(sup 2) and lt; 2.5 at 17% optical-to-optical conversion efficiency. These q-switched results were obtained at a 10 kHz repetition rate and resulted in 77 nsec pulse durations. These improved levels of operational performance have been achieved as a result of technology advancements made in several areas that will be covered in this manuscript. These enhancements to our architecture include: (1) Hollow lens ducts that enable the use of advanced cavity architectures permitting birefringence compensation and the ability to run in large aperture-filling near-diffraction-limited modes. (2) Compound laser rods with flanged-nonabsorbing-endcaps fabricated by diffusion bonding. (3) Techniques for suppressing amplified spontaneous emission (ASE) and parasitics in the polished barrel rods

LASER CUTTING MACHINES FOR 3-D THIN SHEET PARTS

Directory of Open Access Journals (Sweden)

Miroslav RADOVANOVIC

2012-11-01

Full Text Available Laser cutting machines are used for precise contour cutting thin sheet. In industrial application nowadays various types and construction of laser cutting machines can be met. For contour cutting 3-D thin sheet parts laser cutting machines with rotation movements and laser robots are used. Laser generates the light beam, that presents a tool in working process. Application of laser cutting machines made possible good quality of products, flexibility of production and enlargement of economy

Laser Ultrasound Spectroscopy Scanning for 3D Printed Parts

Energy Technology Data Exchange (ETDEWEB)

Brennan, Guendalyn Kendra [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-08-04

One of the challenges of additive manufacturing is quality control due to the possibility of unseen flaws in the final product. The current methods of inspection are lacking in detail, too slow for practical use, or unable to validate internal structure. This report examines the use of laser ultrasound spectroscopy in layer by layer scans of 3D printed parts as they are created. The result is fast and detailed quality control. An additional advantage of this method is the ability to cancel a print as soon as a defect is detected, therefore saving materials and time. This technique, though simple in concept, has been a challenge to implement. I discuss tweaking the 3D printer configuration, and finding the optimal settings for laser scanning small parts made of ABS plastic, as well as the limits of how small of a detail the laser can detect. These settings include the frequency of the ultrasonic transducer, the speed of the laser, and the distance from the laser to the part.

On the Wave Stresses in the Rods of Anvil Hammers

Directory of Open Access Journals (Sweden)

V. M. Sinitskiy

2014-01-01

Full Text Available With operating anvil hammers, there are rigid impacts of die tools, and as a result, almost instantaneous impact stops of the falling parts of hammer. Such operating conditions lead to the accelerated breakdowns of rods because of significant wave stresses arising in them. Common differential and integral methods to estimate wave stresses are widespread in engineering practice. However, to use them a researcher has to possess certain skills and special software. We consider the method for estimating the wave stresses in the rods of anvil hammers based on Laplace transforms (LT of wave equation. The article shows a procedure to set up and solve differential wave equations by operator method. These equations describe the wave propagation process of strains and stresses in the rods of anvil hammers with rigid impact and taking into account a damping rod connection with the head of hammer. The method takes into consideration an influence of both piston and rod weights and of mechanical and geometrical characteristics of rod on the stress value in the placement of rod in hammer head. Results analysis shows that a sufficiently efficient method for practical improving the durability of rods is the method of damping impact load on the rod through setting the damping devices in the form either of elastic "pad" of one or another design or of hydraulic shock absorbers in the placement of its connection with the hammer head. In this case there is a change of the wave front, it becomes flatter. It is shown that the stresses in the rod are proportional to the amount of wave stresses because of the own impact of rod and piston, which make a total weight of the system. Effect of piston weight on the stresses value at the rod during impact is directly proportional to the ratio of its weight to the rod weight. The geometric parameters of rod and the speed of the falling parts before the impact also influence on the value of stresses in the rod.The represented

Cladding For Transversely-Pumped Laser Rod

Science.gov (United States)

Byer, Robert L.; Fan, Tso Yee

1989-01-01

Combination of suitable dimensioning and cladding of neodymium:yttrium aluminum garnet of similar solid-state laser provides for more efficient utilization of transversely-incident pump light from diode lasers. New design overcomes some of limitations of longitudinal- and older transverse-pumping concepts and promotes operation at higher output powers in TEM00 mode.

Laser beam welding of titanium additive manufactured parts

NARCIS (Netherlands)

Wits, Wessel Willems; Jauregui Becker, Juan Manuel

2015-01-01

In this paper the joinability of titanium Additive Manufactured (AM) parts is explored. Keyhole welding, using a pulsed laser beam, of conventionally produced parts is compared to AM parts. Metal AM parts are notorious for having remaining porosities and other non-isotropic properties due to the

Diode laser in-band pumped, efficient 1645 nm continuous-wave and Q-switched Er:YLuAG lasers with near-diffraction-limited beam quality

International Nuclear Information System (INIS)

Li, Jing; Yang, SuHui; He, Tao

2014-01-01

Fiber-like Er:YLuAG laser rods were tested for continuous-wave (CW) and Q-switched operation. Two narrow-band laser diodes emitting at 1532 nm were used as pump sources. The pump power was confined in the laser rods via total internal reflection. In CW mode, a maximum output power of 7.2 W was measured from a 30 mm long Er:YLuAG laser rod, corresponding to an optical–optical efficiency of 26% and a slope efficiency of 78%. Er:YLuAG and Er:YAG lasers were compared experimentally and exhibited comparable performance, while the measured central wavelength of the Er:YLuAG laser was 1644.75 nm, slightly longer than the central wavelength of the Er:YAG laser in the same experimental circumstances. In Q-switched mode, an output energy of 3.5 mJ was obtained from a 25 mm Er:YLuAG laser rod with a pulse duration of 100 ns and a pulse repetition frequency of 100 Hz. The pulsed output had near-diffraction-limited beam quality with M 2 values of 1.13 and 1.11 in the x and y directions, respectively. (letter)

Potential of solar-simulator-pumped alexandrite lasers

Science.gov (United States)

Deyoung, Russell J.

1990-01-01

An attempt was made to pump an alexandrite laser rod using a Tamarak solar simulator and also a tungsten-halogen lamp. A very low optical laser cavity was used to achieve the threshold minimum pumping-power requirement. Lasing was not achieved. The laser threshold optical-power requirement was calculated to be approximately 626 W/sq cm for a gain length of 7.6 cm, whereas the Tamarak simulator produces 1150 W/sq cm over a gain length of 3.3 cm, which is less than the 1442 W/sq cm required to reach laser threshold. The rod was optically pulsed with 200 msec pulses, which allowed the alexandrite rod to operate at near room temperature. The optical intensity-gain-length product to achieve laser threshold should be approximately 35,244 solar constants-cm. In the present setup, this product was 28,111 solar constants-cm.

Control rod

International Nuclear Information System (INIS)

Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

1979-01-01

Purpose: A control rod has been developed, which provided a plurality of through-holes in the vicinity of the sheath fitting position, in order to flatten burn-up, of fuel rods in positions confronting a control rod. Thereby to facilitate the manufacture of the control rods and prevent fuel rod failures. Constitution: A plurality of through-holes are formed in the vicinity of the sheath fitting position of a central support rod to which a sheath for the control rod is fitted. These through-holes are arranged in the axial direction of the central support rod. Accordingly, burn-up of fuel rods confronting the control rods can be reduced by through-holes and fuel rod failures can be prevented. (Yoshino, Y.)

Preliminary nuclear design for test MOX Fuel rods

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

1997-10-01

As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

Guiding and amplification properties of rod-type photonic crystal fibers with sectioned core doping

DEFF Research Database (Denmark)

Selleri, Stefano; Poli, Federica; Passaro, Davide

2009-01-01

Rod-type photonic crystal fibers are large mode area double-cladding fibers with an outer diameter of few millimeters which can provide important advantages for high-power lasers and amplifiers. Numerical studies have recently demonstrated the guidance of higher-order modes in these fibers, which...... can worsen the output beam quality of lasers and amplifiers. In the present analysis a sectioned core doping has been proposed for Yb-doped rod-type photonic crystal fibers, with the aim to improve the higher-order mode suppression. A full-vector modal solver based on the finite element method has...

Highly efficient solar-pumped Nd:YAG laser.

Science.gov (United States)

Liang, Dawei; Almeida, Joana

2011-12-19

The recent progress in solar-pumped laser with Fresnel lens and Cr:Nd:YAG ceramic medium has revitalized solar laser researches, revealing a promising future for renewable reduction of magnesium from magnesium oxide. Here we show a big advance in solar laser collection efficiency by utilizing an economical Fresnel lens and a most widely used Nd:YAG single-crystal rod. The incoming solar radiation from the sun is focused by a 0.9 m diameter Fresnel lens. A dielectric totally internally reflecting secondary concentrator is employed to couple the concentrated solar radiation from the focal zone to a 4 mm diameter Nd:YAG rod within a conical pumping cavity. 12.3 W cw laser power is produced, corresponding to 19.3 W/m(2) collection efficiency, which is 2.9 times larger than the previous results with Nd:YAG single-crystal medium. Record-high slope efficiency of 3.9% is also registered. Laser beam quality is considerably improved by pumping a 3 mm diameter Nd:YAG rod.

Compact and highly efficient laser pump cavity

Science.gov (United States)

Chang, Jim J.; Bass, Isaac L.; Zapata, Luis E.

1999-01-01

A new, compact, side-pumped laser pump cavity design which uses non-conventional optics for injection of laser-diode light into a laser pump chamber includes a plurality of elongated light concentration channels. In one embodiment, the light concentration channels are compound parabolic concentrators (CPC) which have very small exit apertures so that light will not escape from the pumping chamber and will be multiply reflected through the laser rod. This new design effectively traps the pump radiation inside the pump chamber that encloses the laser rod. It enables more uniform laser pumping and highly effective recycle of pump radiation, leading to significantly improved laser performance. This new design also effectively widens the acceptable radiation wavelength of the diodes, resulting in a more reliable laser performance with lower cost.

Seismic analysis of control and safety rod drive mechanism

International Nuclear Information System (INIS)

Meher Prasad, A.; Jaya, K.P.; Chellapandi, P.; Rajan Babu, V.; Selvaraj, T.

2003-01-01

Control rod and its driving mechanism for a Fast Breeder Reactor is to facilitate safe shutdown of the reactor in case of emergency. A theoretical study on the seismic qualification of control and safety rod driving mechanism is carried out. Earthquake excitations under Operational Basis (ORE) and Safe Shutdown condition (SSE) are considered. The time required for the control rod to reach the bottom position in order to shut down the reaction under excited condition is traced out. The maximum displaced positions and extreme stresses in various parts of the system under excitations are evaluated. The system modeled using beam elements. The connections between different parts are modeled through rigid elements. The interaction between various parts are modeled using GAP elements. (author)

Performance of the NRX shut-off rods

International Nuclear Information System (INIS)

Manson, R.E.

1965-08-01

A new type of shut-off rod of electromechanical design was developed by the American Machine and Foundry Company for use in the NRX reactor following the accident of 1952. The new rods were installed in May, 1956, as part of the control system conversion program which was completed in 1958. Some problems were encountered with limit switch adjustment but minor modifications in design led to much improved operation. he performance of the rods also improved as more experience was gained in the maintenance and adjustment of the various headgear components. Each headgear is now overhauled once a year on a routine basis. The present design of shut-off rod is considered to be very satisfactory. There has only been one occasion when a shut-off rod has failed to come fully down on a trip. Rods have failed to operate correctly on five other occasions but these occurred during shutdown periods or when the reactor was being shutdown manually. (author)

Modeling and simulation performance of sucker rod beam pump

Energy Technology Data Exchange (ETDEWEB)

Aditsania, Annisa, E-mail: annisaaditsania@gmail.com [Department of Computational Sciences, Institut Teknologi Bandung (Indonesia); Rahmawati, Silvy Dewi, E-mail: silvyarahmawati@gmail.com; Sukarno, Pudjo, E-mail: psukarno@gmail.com [Department of Petroleum Engineering, Institut Teknologi Bandung (Indonesia); Soewono, Edy, E-mail: esoewono@math.itb.ac.id [Department of Mathematics, Institut Teknologi Bandung (Indonesia)

2015-09-30

Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research.

Modeling and simulation performance of sucker rod beam pump

International Nuclear Information System (INIS)

Aditsania, Annisa; Rahmawati, Silvy Dewi; Sukarno, Pudjo; Soewono, Edy

2015-01-01

Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research

Adjustment of Part Properties for an Elastomeric Laser Sintering Material

Science.gov (United States)

Wegner, A.; Ünlü, T.

2018-03-01

Laser sintering of polymers is gaining more and more importance within the field of small series productions. Polyamide 12 is predominantly used, although a variety of other materials are also available for the laser sintering process. For example, elastomeric, rubberlike materials offer very different part property profiles. Those make the production of flexible parts like, e.g., sealings, flexible tubes or shoe soles possible because they offer high part ductility and low hardness. At the chair for manufacturing technology, a new elastomeric laser sintering material has been developed and then commercialized by a spin-off from university. The aim of the presented study was the analysis of the new material's properties. Proof was found that Shore hardness can be modified by varying the parameter settings. Therefore, the correlation between process parameters, energy input, Shore hardness and other part properties like mechanical properties were analyzed. Based on these results, suitable parameter settings were established which lead to the possibility of producing parts with different Shore hardnesses.

Method of targeted delivery of laser beam to isolated retinal rods by fiber optics.

Science.gov (United States)

Sim, Nigel; Bessarab, Dmitri; Jones, C Michael; Krivitsky, Leonid

2011-11-01

A method of controllable light delivery to retinal rod cells using an optical fiber is described. Photo-induced current of the living rod cells was measured with the suction electrode technique. The approach was tested with measurements relating the spatial distribution of the light intensity to photo-induced current. In addition, the ion current responses of rod cells to polarized light at two different orientation geometries of the cells were studied.

Simulated experimental research on flow field near control rod guide tubes

International Nuclear Information System (INIS)

Yu Ping'an; Shen Xiuzhong; Yang Guanyue; He Fangzheng; Gao Weiguo; Zhang Zhiyi; Tian Ji'an

1997-01-01

The paper presents the velocity measurement in the 1/4 scale transparent model of PWR pressure vessel upper plenum of 300 MW nuclear power plant by employing dynamic resistance strain foil velocity measurement technology and laser Doppler velocity measurement technology which have no effect on the flow field. In the experiment water is chosen as the fluid. As a result of the measurement the hydraulic load on the control rods is clarified and the experimental basis is provided for the analysis of whether the control rods are moving upward and downward freely and drop rapidly in emergency case by order. Meantime it also provides the experimental basis for the optical design of the control rod guide tubes and bundles

Analysis of buffering process of control rod hydraulic absorber

International Nuclear Information System (INIS)

Bao Jishi; Qin Benke; Bo Hanliang

2011-01-01

Control Rod Hydraulic Drive Mechanism(CRHDM) is a newly invented build-in control rod drive mechanism. Hydraulic absorber is the key part of this mechanism, and is used to cushion the control rod when the rod scrams. Thus, it prevents the control rod from being deformed and damaged. In this paper dynamics program ANSYS CFX is used to calculate all kinds of flow conditions in hydraulic absorber to obtain its hydraulic characteristics. Based on the flow resistance coefficients obtained from the simulation results, fluid mass and momentum equations were developed to get the trend of pressure change in the hydraulic cylinder and the displacement of the piston rod during the buffering process of the control rod. The results obtained in this paper indicate that the hydraulic absorber meets the design requirement. The work in this paper will be helpful for the design and optimization of the control rod hydraulic absorber. (author)

Control rod drive mechanism

International Nuclear Information System (INIS)

Mizuno, Katsuyuki.

1976-01-01

Object: To restrict the reduction in performance due to stress corrosion cracks by making use of condensate produced in a turbine steam condenser. Structure: Water produced in a turbine steam condenser is forced into a condensed water desalting unit by low pressure condensate pump. The condensate is purified and then forced by a high pressure condensate pump into a feedwater heater for heating before it is returned to the reactor by a feedwater pump. Part of the condensate issuing from the condensate desalting unit is branched from the remaining portion at a point upstream the pump and is withdrawn into a control rod drive water pump after passing through a motordriven bypass valve, an orifice and a condenser water level control valve, is pressurized in the control rod drive water desalting unit and supplied to a control rod drive water pressure system. The control rod is vertically moved by the valve operation of the water pressure system. Since water of high oxygen concentration does not enter during normal operation, it is possible to prevent the stress cracking of the stainless steel apparatus. (Nakamura, S.)

Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

2014-02-15

Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

Control rod

International Nuclear Information System (INIS)

Igarashi, Takao; Sugawara, Satoshi; Yoshimoto, Yuichiro; Saito, Shozo; Fukumoto, Takashi.

1987-01-01

Purpose: To reduce the weight and thereby obtain satisfactory operationability of control rods by combining absorbing nuclear chain type neutron absorbers and conventional type neutron absorbers in the axial direction of blades. Constitution: Neutron absorber rods and long life type neutron absorber rods are disposed in a tie rod and a sheath. The neutron absorber rod comprises a poison tube made of stainless steels and packed with B 4 C powder. The long life type neutron absorber rod is prepared by packing B-10 enriched boron carbide powder into a hafnium metal rod, hafnium pipe, europium and stainless made poison tube. Since the long life type absorber rod uses HF as the absorbing nuclear chain type neutron absorber, it absorbs neutrons to form new neutron absorbers to increase the nuclear life. (Yoshino, Y.)

Dynamic rod worth measurements (''Rod Insertion''). Final report for the period 01 December 1994 - 30 November 1996

International Nuclear Information System (INIS)

Bogdan, G.

1996-12-01

Reload startup physics tests are performed for pressurized water reactors (PWR power plant) following a refuelling or other significant core alteration for which nuclear design calculations are required. Part of the reload startup physics tests are control rod group worths measurements. for this purpose a new so-called method ''Rod-Insertion'' was developed. It can also be used as an additional measuring instrument on the research reactor for education purposes. The principle of the rod-insertion method is to start from a critical reactor operating at low power and to measure the time-dependent reactivity change while a control rod is inserted into the core. Unlike in the rod-drop method, the measured control rod is inserted with the drive mechanism at normal speed. By analyzing the flux trace using point-kinetics, not only the total rod worth but also the differential and the integral rod worth curves are obtained. A high-quality electrometer is required for monitoring the neutron flux. The analysis is performed by transferring the data to an IBM PC compatible with some additional standard electronic board and the associated software. The new reactivity meter has been validated on the TRIGA Mark II reactors in Ljubljana and Vienna and at the Krsko Nuclear Power Plant during physics startup tests after reload. The results proved the high performance of the reactivity meter in the standard applications according to the existing procedures, as well as in the new rod-insertion technique of measuring the control rod group worths. This method drastically differs from others such as absence of any chemical control of reactivity (like boron exchange method), and minimizing a testing time and waste coolant production

Guiding and amplification properties of rod-type photonic crystal fibers with sectioned core doping

Science.gov (United States)

Selleri, S.; Poli, F.; Passaro, D.; Cucinotta, A.; Lægsgaard, J.; Broeng, J.

2009-05-01

Rod-type photonic crystal fibers are large mode area double-cladding fibers with an outer diameter of few millimeters which can provide important advantages for high-power lasers and amplifiers. Numerical studies have recently demonstrated the guidance of higher-order modes in these fibers, which can worsen the output beam quality of lasers and amplifiers. In the present analysis a sectioned core doping has been proposed for Ybdoped rod-type photonic crystal fibers, with the aim to improve the higher-order mode suppression. A full-vector modal solver based on the finite element method has been applied to properly design the low refractive index ring in the fiber core, which can provide an increase of the differential overlap between the fundamental and the higher-order mode. Then, the gain competition among the guided modes along the Yb-doped rod-type fibers has been investigated with a spatial and spectral amplifier model. Simulation results have shown the effectiveness of the sectioned core doping in worsening the higher-order mode overlap on the doped area, thus providing an effective single-mode behavior of the Yb-doped rod-type photonic crystal fibers.

Control rods

International Nuclear Information System (INIS)

Maruyama, Hiromi.

1984-01-01

Purpose: To realize effective utilization, cost reduction and weight reduction in neutron absorbing materials. Constitution: Residual amount of neutron absorbing material is averaged between the top end region and other regions of a control rod upon reaching to the control rod working life, by using a single kind of neutron absorbing material and increasing the amount of the neutron absorber material at the top end region of the control rod as compared with that in the other regions. Further, in a case of a control rod having control rod blades such as in a cross-like control rod, the amount of the neutron absorbing material is decreased in the middle portion than in the both end portions of the control rod blade along the transversal direction of the rod, so that the residual amount of the neutron absorbing material is balanced between the central region and both end regions upon reaching the working life of the control rod. (Yoshihara, H.)

Inspection of additive manufactured parts using laser ultrasonics

Science.gov (United States)

Lévesque, D.; Bescond, C.; Lord, M.; Cao, X.; Wanjara, P.; Monchalin, J.-P.

2016-02-01

Additive manufacturing is a novel technology of high importance for global sustainability of resources. As additive manufacturing involves typically layer-by-layer fusion of the feedstock (wire or powder), an important characteristic of the fabricated metallic structural parts, such as those used in aero-engines, is the performance, which is highly related to the presence of defects, such as cracks, lack of fusion or bonding between layers, and porosity. For this purpose, laser ultrasonics is very attractive due to its non-contact nature and is especially suited for the analysis of parts of complex geometries. In addition, the technique is well adapted to online implementation and real-time measurement during the manufacturing process. The inspection can be performed from either the top deposited layer or the underside of the substrate and the defects can be visualized using laser ultrasonics combined with the synthetic aperture focusing technique (SAFT). In this work, a variety of results obtained off-line on INCONEL® 718 and Ti-6Al-4V coupons that were manufactured using laser powder, laser wire, or electron beam wire deposition are reported and most defects detected were further confirmed by X-ray micro-computed tomography.

Laser doppler and radar interferometer for contactless measurements on unaccessible tie-rods on monumental buildings: Santa Maria della Consolazione Temple in Todi

International Nuclear Information System (INIS)

Gioffré, M; Cavalagli, N; Pepi, C; Trequattrini, M

2017-01-01

Non-contact measurements can be effectively used in civil engineering to assess the variation of structural performance with time. In the last decades this approach has received considerable interests from researchers working in the field of structural health monitoring (SHM). Indeed, non-contact measurements are very attractive because it is possible to perform non intrusive and non destructive investigations even being at a significant distance from the targets. Within this context, contactless measurements of the tie-rod vibrations in the Santa Maria della Consolazione Temple in Todi (Italy) are presented in this paper. In particular, laser vibrometer and radar interferometer measurements are used to estimate natural frequencies and mode shapes. This information is crucial to obtain the tensile axial force in the tie-rods, which can be used as an indicator of structural integrity or possible failure. Furthermore, a novel approach is proposed where drones (Unmanned Aerial Vehicles) can be successfully used to improve the effectiveness and the accuracy of the experimental activities. (paper)

Control rod calibration including the rod coupling effect

International Nuclear Information System (INIS)

Szilard, R.; Nelson, G.W.

1984-01-01

In a reactor containing more than one control rod, which includes all reactors licensed in the United States, there will be a 'coupling' or 'shadowing' of control rod flux at the location of a control rod as a result of the flux depression caused by another control rod. It was decided to investigate this phenomenon further, and eventually to put calibration table data or formulae in a small computer in the control room, so once could insert the positions of the three control rods and receive the excess reactivity without referring to separate tables. For this to be accomplished, a 'three control- rod reactivity function' would be used which would include the flux coupling between the rods. The function is design and measured data was fitted into it to determine the calibration constants. The input data for fitting the trial functions consisted of 254 data points, each consisting of the position of the reg, shim, and transient rods, and the total excess reactivity. (About 200 of these points were 'critical balance points', that is the rod positions for which reactor was critical, and the remainder were determined by positive period measurements.) Although this may be unrealistic from a physical viewpoint, the function derived gave a very accurate recalculation of the input data, and thus would faithfully give the excess reactivity for any possible combination of the locations of the three control rods. The next step, incorporation of the three-rod function into the minicomputer, will be pursued in the summer and fall of 1984

Laser beam welding of titanium additive manufactured parts

OpenAIRE

Wits, Wessel Willems; Jauregui Becker, Juan Manuel

2015-01-01

In this paper the joinability of titanium Additive Manufactured (AM) parts is explored. Keyhole welding, using a pulsed laser beam, of conventionally produced parts is compared to AM parts. Metal AM parts are notorious for having remaining porosities and other non-isotropic properties due to the layered manufacturing process. This study shows that due to these deficiencies more energy per unit weld length is required to obtain a similar keyhole geometry for titanium AM parts. It is also demon...

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 1 of Volume IV, discusses: Process overview functional descriptions; Control system descriptions; Support system descriptions; Maintenance system descriptions; and Process equipment descriptions

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 4 of Volume IV, discusses: Off-normal operating and recovery procedures; Emergency response procedures; Troubleshooting procedures; and Preventive maintenance procedures

The estimation of the control rods absorber burn-up during the VVER-1000 operation

Energy Technology Data Exchange (ETDEWEB)

Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

2013-09-15

The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

Cadmium safety rod thermal tests

International Nuclear Information System (INIS)

Thomas, J.K.; Iyer, N.C.; Peacock, H.B.

1992-01-01

Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing

Study on process of laser cladded nuclear valve parts

International Nuclear Information System (INIS)

Zhang Chunliang

2000-01-01

The microstructure and performances of the Co-base alloy coatings that are formed by laser cladding, plasma spurt welding and arc surfacing on the nuclear valve-sealing surface have been studied and compared. The combination costs of laser cladding, plasma spurt welding and arc, surfacing have been analyzed and compared. The results showed that the laser cladding processing has the advantages of high efficiency, low energy cost, a little machining allowance, high rate of finished products and low combination cost, compared with plasma spurt welding processing and arc surfacing processing. The laser cladding technology can improve the qualities of nuclear valve parts and increase their service life. Therefore, the laser cladding processing is a new technology with developing potential

Method of driving control rod in reactor

International Nuclear Information System (INIS)

Osa, Hirotaka.

1986-01-01

Purpose: To improve security and safety of the reactor by reducing reactor output automatically and quickly when circulation of cooling water is stopped. Constitution: When the circulating pump is under operation, fluid pressure in the discharge pipe is transferred to the fluid room of fluid pressure cylinder via the control rod drive pipe and lift up the piston, and then the control rod is drawn out of the reactor core. When the circulating pump is lowered in its functions, discharge pipe fluid pressure decreases, fluid pressure in the fluid room decreases, and with less force of piston movement, the control rod gets lowered by its own weight. At this time, the blocked state of the opening by the piston is released, fluid flows into the room. Lowering of pressure and the control rod is promoted by transferring out fluid below the piston in the fluid room to the upper part of the piston via a small gap when the control rod falls by gravity. (Horiuchi, T.)

Time-dependent analytical thermal model to investigate thermally induced stresses in quasi-CW-pumped laser rods

CSIR Research Space (South Africa)

Bernhardi, EH

2008-01-01

Full Text Available that determines the temperature and the thermally induced stresses in isotropic rods is presented. Even though the model is developed for isotropic rods, it is shown that it can also be used to accurately estimate the thermal effects in anisotropic rods...

Development of a non-piston MR suspension rod for variable mass systems

Science.gov (United States)

Deng, Huaxia; Han, Guanghui; Zhang, Jin; Wang, Mingxian; Ma, Mengchao; Zhong, Xiang; Yu, Liandong

2018-06-01

The semi-active suspension systems for variable mass systems require long work stroke and variable damping, while the currently piston structure limits the work stroke for the magnetorheological (MR) dampers. The main work of this paper is to design a semi-active non-piston MR (NPMR) suspension rod for the reduction of the vibration of an automatic impeller washing machine, which is a typical variable mass system. The designed suspension rod locates in the suspension system that links the internal tub to the washing machine cabinet. The NPMR suspension rod includes a MR part and a air part. The MR part can provide low initial damping force and the unlimited work stroke compared with the piston MR damper. The hysteretic response tests and vibration performance evaluation with different loadings are conducted to verify the dynamic performance for the designed rod. The measured damping force of the MR part varies from 5 to 20 N. Studies of dehydration mode experiments of the washing machine indicate that its vibration acceleration with the NPMR suspension rods can reduce to half of the original passive ones in certain conditions.

Control rod drives

International Nuclear Information System (INIS)

Nakamura, Akira.

1984-01-01

Purpose: To enable to monitor the coupling state between a control rod and a control rod drive. Constitution: After the completion of a control rod withdrawal, a coolant pressure is applied to a control rod drive being adjusted so as to raise only the control rod drive and, in a case where the coupling between the control rod drive and the control rod is detached, the former is elevated till it contacts the control rod and then stopped. The actual stopping position is detected by an actual position detection circuit and compared with a predetermined position stored in a predetermined position detection circuit. If both of the positions are not aligned with each other, it is judged by a judging circuit that the control rod and the control rod drives are not combined. (Sekiya, K.)

A nuclear reactor with buffered control rods

International Nuclear Information System (INIS)

Bevilacqua, F.

1974-01-01

The control rods for, e.g., water-cooled reactors are fastened as units on common crossbars in vertical downward direction. The fastening on the crossbar is achieved by means of cross-shaped parts, e.g., in the shape of a double 'H'. A cylinder connected with a drive rod in normal operation is joined to each of the crossbars. In an emergency shut-down, this connection is interrupted and the control rod unit drops into the core through the action of gravity. Its fall is slowed down by a cushion or shock absorbing unit. For this purpose a piston is provided mounted on the supporting plate below the cylinder and guided within it. In the cylinder, the coolant is contained as damping medium. An upper opening in the cylinder serves as a ventilation hole. The movement of the piston is limited by a stopping part within the cylinder and slowed down by a spiral spring. (DG) [de

Lasers with intra-cavity phase elements

Science.gov (United States)

Gulses, A. Alkan; Kurtz, Russell; Islas, Gabriel; Anisimov, Igor

2018-02-01

Conventional laser resonators yield multimodal output, especially at high powers and short cavity lengths. Since highorder modes exhibit large divergence, it is desirable to suppress them to improve laser quality. Traditionally, such modal discriminations can be achieved by simple apertures that provide absorptive loss for large diameter modes, while allowing the lower orders, such as the fundamental Gaussian, to pass through. However, modal discrimination may not be sufficient for short-cavity lasers, resulting in multimodal operation as well as power loss and overheating in the absorptive part of the aperture. In research to improve laser mode control with minimal energy loss, systematic experiments have been executed using phase-only elements. These were composed of an intra-cavity step function and a diffractive out-coupler made of a computer-generated hologram. The platform was a 15-cm long solid-state laser that employs a neodymium-doped yttrium orthovanadate crystal rod, producing 1064 nm multimodal laser output. The intra-cavity phase elements (PEs) were shown to be highly effective in obtaining beams with reduced M-squared values and increased output powers, yielding improved values of radiance. The utilization of more sophisticated diffractive elements is promising for more difficult laser systems.

Design techniques and measured performance for a uniformly-pumped 4-cm diameter rod amplifier

International Nuclear Information System (INIS)

Linford, G.J.; Yarema, S.M.

1976-01-01

A solid-state laser rod amplifier of moderate aperture achieving a high degree of spatial gain uniformity has been constructed and its performance evaluated. Digital and analogue techniques were used to optimize the amplifier design for performance in a laser fusion application. Results of simple 2-D computer simulations and experimental evaluations of amplifier performance are presented

Part I: $\\beta$-delayed fission, laser spectroscopy and shape-coexistence studies with astatine beams; Part II: Delineating the island of deformation in the light gold isotopes by means of laser spectroscopy

CERN Document Server

Andreyev, Andrei

2013-01-01

Part I: $\\beta$-delayed fission, laser spectroscopy and shape-coexistence studies with astatine beams; Part II: Delineating the island of deformation in the light gold isotopes by means of laser spectroscopy

Device for coupling a control rod and control rod drive

International Nuclear Information System (INIS)

Nishioka, Kazuya.

1975-01-01

Object: To obtain simple and reliable coupling between a control rod and control rod drive by equipping the lower end of the control rod with an extension provided with lateral protuberances and forming the upper end of an index tube with a recess provided with lateral holes. Structure: The tapering central extension of the control rod is inserted into the recess by lowering the control rod, and then it is further inserted by causing frictional movement of the inclined surfaces of lateral protuberances in frictional contact with guide surfaces. When the lateral protuberances are brought into contact with a stepped portion, the control rod is rotated to fit the lateral protuberances into the lateral holes. In this way, the control rod is coupled to the index tube of the control rod drive. (Yoshino, Y.)

Radioactive lightning rods waste treatment

International Nuclear Information System (INIS)

Vicente, Roberto; Dellamano, Jose C.; Hiromoto, Goro

2008-01-01

Full text: In this paper, we present alternative processes that could be adopted for the management of radioactive waste that arises from the replacement of lightning rods with attached Americium-241 sources. Lightning protectors, with Americium-241 sources attached to the air terminals, were manufactured in Brazil until 1989, when the regulatory authority overthrew the license for fabrication, commerce, and installation of radioactive lightning rods. It is estimated that, during the license period, about 75,000 such devices were set up in public, commercial and industrial buildings, including houses and schools. However, the policy of CNEN in regard to the replacement of the installed radioactive rods, has been to leave the decision to municipal governments under local building regulations, requiring only that the replaced rods be sent immediately to one of its research institutes to be treated as radioactive waste. As a consequence, the program of replacement proceeds in a low pace and until now only about twenty thousand rods have reached the waste treatment facilities The process of management that was adopted is based primarily on the assumption that the Am-241 sources will be disposed of as radioactive sealed sources, probably in a deep borehole repository. The process can be described broadly by the following steps: a) Receive and put the lightning rods in initial storage; b) Disassemble the rods and pull out the sources; c) Decontaminate and release the metal parts to metal recycling; d) Store the sources in intermediate storage; e) Package the sources in final disposal packages; and f) Send the sources for final disposal. Up to now, the disassembled devices gave rise to about 90,000 sources which are kept in storage while the design of the final disposal package is in progress. (author)

Control rod displacement

International Nuclear Information System (INIS)

Nakazato, S.

1987-01-01

This patent describes a nuclear reactor including a core, cylindrical control rods, a single support means supporting the control rods from their upper ends in spaced apart positions and movable for displacing the control rods in their longitudinal direction between a first end position in which the control rods are fully inserted into the core and a second end position in which the control rods are retracted from the core, and guide means contacting discrete regions of the outer surface of each control rod at least when the control rods are in the vicinity of the second end position. The control rods are supported by the support means for longitudinal movement without rotation into and out of the core relative to the guide means to thereby cause the outer surface of the control rods to experience wear as a result of sliding contact with the guide means. The support means are so arranged with respect to the core and the guide means that it is incapable of rotation relative to the guide means. The improvement comprises displacement means being operatively coupled to a respective one of the control rods for periodically rotating the control rod in a single angular direction through an angle selected to change the locations on the outer surfaces of the control rods at which the control rods are contacted by the guide means during subsequent longitudinal movement of the control rods

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 2 discusses the following topics: Fuel Rod Extraction System Test Results and Analysis Reports and Clamping Table Test Results and Analysis Reports

High power diode pumped solid state (DPSS) laser systems active media robust modeling and analysis

Science.gov (United States)

Kashef, Tamer M.; Mokhtar, Ayman M.; Ghoniemy, Samy A.

2018-02-01

Diode side-pumped solid-state lasers have the potential to yield high quality laser beams with high efficiency and reliability. This paper summarizes the results of simulation of the most predominant active media that are used in high power diode pumped solid-state (DPSS) laser systems. Nd:YAG, Nd:glass, and Nd:YLF rods laser systems were simulated using the special finite element analysis software program LASCAD. A performance trade off analysis for Nd:YAG, Nd:glass, and Nd:YLF rods was performed in order to predict the system optimized parameters and to investigate thermally induced thermal fracture that may occur due to heat load and mechanical stress. The simulation results showed that at the optimized values Nd:YAG rod achieved the highest output power of 175W with 43% efficiency and heat load of 1.873W/mm3. A negligible changes in laser output power, heat load, stress, and temperature distributions were observed when the Nd:YAG rod length was increased from 72 to 80mm. Simulation of Nd:glass at different rod diameters at the same pumping conditions showed better results for mechanical stress and thermal load than that of Nd:YAG and Nd:YLF which makes it very suitable for high power laser applications especially for large rod diameters. For large rod diameters Nd:YLF is mechanically weaker and softer crystal compared to Nd:YAG and Nd:glass due to its poor thermomechanical properties which limits its usage to only low to medium power systems.

Control rod drive

International Nuclear Information System (INIS)

Hawke, B.C.

1986-01-01

A reactor core, one or more control rods, and a control rod drive are described for selectively inserting and withdrawing the one or more control rods into and from the reactor core, which consists of: a support structure secured beneath the reactor core; control rod positioning means supported by the support structure for movably supporting the control rod for movement between a lower position wherein the control rod is located substantially beneath the reactor core and an upper position wherein at least an upper portion of the control rod extends into the reactor core; transmission means; primary drive means connected with the control rod positioning means by the transmission means for positioning the control rod under normal operating conditions; emergency drive means for moving the control rod from the lower position to the upper position under emergency conditions, the emergency drive means including a weight movable between an upper and a lower position, means for movably supporting the weight, and means for transmitting gravitational force exerted on the weight to the control rod positioning means to move the control rod upwardly when the weight is pulled downwardly by gravity; the transmission means connecting the control rod positioning means with the emergency drive means so that the primary drive means effects movement of the weight and the control rod in opposite directions under normal conditions, thus providing counterbalancing to reduce the force required for upward movement of the control rod under normal conditions; and restraint means for restraining the fall of the weight under normal operating conditions and disengaging the primary drive means to release the weight under emergency conditions

A comparison of thermal algorithms of fuel rod performance code systems

International Nuclear Information System (INIS)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

A comparison of thermal algorithms of fuel rod performance code systems

Energy Technology Data Exchange (ETDEWEB)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

French LMFBR's control rods experience and development

International Nuclear Information System (INIS)

Arnaud, G.; Guigon, A.; Verset, L.

1983-06-01

Since the last ten years, the French program has been, first of all, directed to the setting up, and then the development of, at once, the Phenix control rods, and next, the Super-Phenix ones. The vented pin design, with porous plug and sodium bonding, which allows the choices of large diameters, has been taken, since the Rapsodie experience was decisive. The absorber material is sintered, 10 B enriched, boron carbide. The can is made of 316 type stainless steel, stabilised, or not, with titanium. The experience gained in Phenix up to now is important, and deals with about six loads of control rods. Results confirm the validity of the design of the absorber pins. Some difficulties has been encountered for the guiding devices, due to the swelling of the steel. They have required design and material improvements. Such difficulties are discarded by a new design of the bearing, for the Super-Phenix control rods. The other parts of these rods, from the Primary Shut-Down System, are strictly derived from Phenix. The design of the rods from the Secondary Shut-Down System is rather different, but it's not the case for the design of the absorber pins: in many a way, they are derived from Phenix pins and from Rapsodie control rods. Both types of rods irradiation tests are in progress in Phenix [fr

Pressure drop ana velocity measurements in KMRR fuel rod bundles

International Nuclear Information System (INIS)

Yagn, Sun Kyu; Chung, Heung June; Chung, Chang Whan; Chun, Se Young; Song, Chul Wha; Won, Soon Yeun; Chung, Moon Ki

1990-01-01

The detailed hydraulic characteristic measurements in subchannels of longitudinally finned rod bundles using one-component LDV(Laser Doppler Velocimeter) were performed. Time mean axial velocity, turbulent intensity, and turbulent micro scales, such as time auto-correlation, Eulerian integral and micro scale, Kolmogorov length and time scale, and Taylor micro length scale were measured. The signals from LDV are inherently more or less discontinuous. The spectra of signals having such intermittent defects can be obtained by the fast Fourier transformation (FFT) of the auto-correlation function. The turbulent crossflow mixing rate between neighboring subchannels and dominant frequencies were evaluated from the measured data. Pressure drop data were obtained for the typical 36-element and 18-element fuel rod bundles fabricated by the design requirement of KMRR fuel and for other type of fuels assembled with 6-fin rods to investigate the fin effects on the pressure drop characteristics

Effects of thermocouple installation and location on fuel rod temperature measurements

International Nuclear Information System (INIS)

McCormick, R.D.

1983-01-01

This paper describes the results of analyses of nuclear fuel rod cladding temperature data obtained during in-reactor experiments under steady state and transient (simulated loss-of-coolant accident) operating conditions. The objective of the analyses was to determine the effect of thermocouple attachment method and location on measured thermal response. The use of external thermocouples increased the time to critical heat flux (CHF), reduced the blowdown peak temperature, and enhanced rod quench. A comparison of laser welded and resistance welded external thermocouple responses showed that the laser welding technique reduced the indicated cladding steady state temperatures and provided shorter time-to-CHF. A comparison of internal welded and embedded thermocouples indicated that the welded technique gave generally unsatisfactory cladding temperature measurements. The embedded thermocouple gave good, consistent results, but was possibly more fragile than the welded thermocouples. Detailed descriptions of the thermocouple designs, attachment methods and locations, and test conditions are provided

Post-processing of 3D-printed parts using femtosecond and picosecond laser radiation

Science.gov (United States)

Mingareev, Ilya; Gehlich, Nils; Bonhoff, Tobias; Meiners, Wilhelm; Kelbassa, Ingomar; Biermann, Tim; Richardson, Martin C.

2014-03-01

Additive manufacturing, also known as 3D-printing, is a near-net shape manufacturing approach, delivering part geometry that can be considerably affected by various process conditions, heat-induced distortions, solidified melt droplets, partially fused powders, and surface modifications induced by the manufacturing tool motion and processing strategy. High-repetition rate femtosecond and picosecond laser radiation was utilized to improve surface quality of metal parts manufactured by laser additive techniques. Different laser scanning approaches were utilized to increase the ablation efficiency and to reduce the surface roughness while preserving the initial part geometry. We studied post-processing of 3D-shaped parts made of Nickel- and Titanium-base alloys by utilizing Selective Laser Melting (SLM) and Laser Metal Deposition (LMD) as additive manufacturing techniques. Process parameters such as the pulse energy, the number of layers and their spatial separation were varied. Surface processing in several layers was necessary to remove the excessive material, such as individual powder particles, and to reduce the average surface roughness from asdeposited 22-45 μm to a few microns. Due to the ultrafast laser-processing regime and the small heat-affected zone induced in materials, this novel integrated manufacturing approach can be used to post-process parts made of thermally and mechanically sensitive materials, and to attain complex designed shapes with micrometer precision.

Sucker rods

Energy Technology Data Exchange (ETDEWEB)

Rylov, B M; Kostur, I N; Shcheigiy, B I; Sukhanov, V S

1983-01-01

As an addendum to A.s. USSR patent No 769087, this particular sucker rod utilizes a differential piston spring that has been attached outside the body of the auxiliary pump. The pump cylinder is attached to the intake line of the main pump. The lower part of the auxiliary pump is equipped with vertical slits, while the differential piston is equipped with a perforated pusher and support under the spring; it can also be shifted as necessary with respect to the vertical slits.

Mode profiling of optical fibers at high laser powers

DEFF Research Database (Denmark)

Nielsen, Peter Carøe; Pedersen, David Bue; Simonsen, R.B.

2008-01-01

of the focused spot can be determined. The analyser is based on the principle of a rotating wire being swept though the laser beam, while the reflected signal is recorded [1]. By changing the incident angle of the rotating rod from 0° to 360° in relation to the fiber, the full profile of the laser beam...... is obtained. Choosing a highly reflective rod material and a sufficiently high rotation speed, these measurements can be done with high laser powers, without any additional optical elements between the fiber and analyzer. The performance of the analyzer was evaluated by coupling laser light into different...

Continuous-wave ceramic Nd:YAG laser at 1123 nm

International Nuclear Information System (INIS)

Zhang, S S; Wang, Q P; Zhang, X Y; Cong, Z H; Fan, S Z; Liu, Z J; Sun, W J

2009-01-01

Ceramic Nd:YAG (cNd:YAG) materials are employed to generate 1123-nm laser. A fiber-coupled continuous-wave (CW) 808-nm diode laser is used as the pumping source. With an incident diode power of 26.1 W, a CW output power of up to 10.8 W is obtained with a 10-mm-long ceramic Nd:YAG rod (1.0 at.%-Nd-doped). The conversion efficiency from diode power to 1123-nm laser power is 41.4%. The laser performance of another 10-mm-long cNd:YAG rod with a Nd-doping concentration of 0.6 at.% is studied as a comparison

Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

International Nuclear Information System (INIS)

Wilkins, S.C.

1977-06-01

Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

Design requirement on KALIMER control rod assembly duct

International Nuclear Information System (INIS)

Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J.

1998-03-01

This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

Design requirement on KALIMER control rod assembly duct

Energy Technology Data Exchange (ETDEWEB)

Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

1998-03-01

This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

Solar-Pumped TEM₀₀ Mode Nd:YAG laser.

Science.gov (United States)

Liang, Dawei; Almeida, Joana

2013-10-21

Here we show a significant advance in solar-pumped laser beam brightness by utilizing a 1.0 m diameter Fresnel lens and a 3 mm diameter Nd:YAG single-crystal rod. The incoming solar radiation is firstly focused by the Fresnel lens on a solar tracker. A large aspheric lens and a 2D-CPC concentrator are then combined to further compress the concentrated solar radiation along the thin laser rod within a V-shaped pumping cavity. 2.3 W cw TEM₀₀ (M² ≤ 1.1) solar laser power is finally produced, attaining 1.9 W laser beam brightness figure of merit, which is 6.6 times higher than the previous record. For multimode operation, 8.1 W cw laser power is produced, corresponding to 143% enhancement in collection efficiency.

External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor

International Nuclear Information System (INIS)

Welty, R.K.

1980-01-01

The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications, and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis

Laser cutting of concretes with various ballasts

International Nuclear Information System (INIS)

Hamasaki, Masanobu; Katsumura, Munehide; Utsumi, Hiroaki

1985-01-01

The biological shield concrete and the radiation shield concrete which construct a part of the atomic reactor must be demolished with the decommissioning of the atomic reactor plants. In a case, the demolition using a laser is expected as one of excellent method for the decommissioning of these radioactive concretes. The fundamental cuttings of a mortar, the concretes with andesite, lime stone and gray wacke as ballast and a concrete reinforced with mild steel rods were therefore carried out using a 5 kW output CO 2 laser. As the results of experiment, it was cleared that cutting results varied with ballasts, 100 mm thick reinforced concrete could be cut, safety was high because few dross and few fume were produced. (author)

Replacement rod

International Nuclear Information System (INIS)

Hatfield, S.C.

1989-01-01

This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

Single-mode amplification in Yb-doped rod-type photonic crystal fibers for high brilliance lasers

DEFF Research Database (Denmark)

Poli, F.; Lægsgaard, Jesper; Passaro, D.

2009-01-01

This paper presents the effect of a low refractive index ring in the Yb-doped rod-type photonic crystal fibre core on the guided mode propagation and analyzed through a spatial and spectral amplifier model. The ring provides a higher differential overlap between the fundamental mode (FM...... to identify a proper ring characteristic that is width, position and refractive index. Then rod-type PCF designs have been optimized with a full-vector modal solver based on the finite-element method. Then, the amplification properties of the Yb-doped rod-type PCFs have been investigated by assuming a forward...

Water rod

International Nuclear Information System (INIS)

Kashiwai, Shin-ichi; Yokomizo, Osamu; Orii, Akihito.

1992-01-01

In a reactor core of a BWR type reactor, the area of a flow channel in a lower portion of a downcoming pipe for downwardly releasing steams present at the top portion in a water rod is increased. Further, a third coolant flow channel (an inner water rod) is disposed in an uprising having an exit opened near the inlet of the water rod and an inlet opened at the outside near the top portion of the water and having an increase flow channel area in the upper portion. The downcoming pipe in the water rod is filled with steams, and the void ratio is increased by so much as the flow channel area of the downcoming pipe is increased. Since the pressure difference between the inlet and the exit of the inner water rod is greater than the pressure difference between the inlet and the exit of the water rod, most of water flown into the inner water rod is discharged out of the exit in the form of water as it is. Since the area of the flow channel is increased in the portion of the inner water rod, void efficiency in the upper portion of the reactor core is decreased by so much. Since the void ratio is thus increased in the lower portion and the void efficiency is decreased in the upper portion of the reactor core, axial void distribution can be flattened. (N.H.)

Oxide nano-rod array structure via a simple metallurgical process

International Nuclear Information System (INIS)

Nanko, M; Do, D T M

2011-01-01

A simple method for fabricating oxide nano-rod array structure via metallurgical process is reported. Some dilute alloys such as Ni(Al) solid solution shows internal oxidation with rod-like oxide precipices during high-temperature oxidation with low oxygen partial pressure. By removing a metal part in internal oxidation zone, oxide nano-rod array structure can be developed on the surface of metallic components. In this report, Al 2 O 3 or NiAl 2 O 4 nano-rod array structures were prepared by using Ni(Al) solid solution. Effects of Cr addition into Ni(Al) solid solution on internal oxidation were also reported. Pack cementation process for aluminizing of Ni surface was applied to prepare nano-rod array components with desired shape. Near-net shape Ni components with oxide nano-rod array structure on their surface can be prepared by using the pack cementation process and internal oxidation,

High-power diode laser bars as pump sources for fiber lasers and amplifiers (Invited Paper)

Science.gov (United States)

Bonati, G.; Hennig, P.; Wolff, D.; Voelckel, H.; Gabler, T.; Krause, U.; T'nnermann, A.; Reich, M.; Limpert, J.; Werner, E.; Liem, A.

2005-04-01

Fiber lasers are pumped by fibercoupled, multimode single chip devices at 915nm. That"s what everybody assumes when asked for the type of fiber laser pumps and it was like this for many years. Coming up as an amplifier for telecom applications, the amount of pump power needed was in the range of several watts. Highest pump powers for a limited market entered the ten watts range. This is a range of power that can be covered by highly reliable multimode chips, that have to survive up to 25 years, e.g. in submarine applications. With fiber lasers entering the power range and the application fields of rod and thin disc lasers, the amount of pump power needed raised into the area of several hundred watts. In this area of pump power, usually bar based pumps are used. This is due to the much higher cost pressure of the industrial customers compared to telecom customers. We expect more then 70% of all industrial systems to be pumped by diode laser bars. Predictions that bar based pumps survive for just a thousand hours in cw-operation and fractions of this if pulsed are wrong. Bar based pumps have to perform on full power for 10.000h on Micro channel heat sinks and 20.000h on passive heatsinks in industrial applications, and they do. We will show a variety of data, "real" long time tests and statistics from the JENOPTIK Laserdiode as well as data of thousands of bars in the field, showing that bar based pumps are not just well suitable for industrial applications on high power levels, but even showing benefits compared to chip based pumps. And it"s reasonable, that the same objectives of cost effectiveness, power and lifetime apply as well to thin disc, rod and slab lasers as to fiber lasers. Due to the pumping of fiber lasers, examples will be shown, how to utilize bars for high brightness fiber coupling. In this area, the automation is on its way to reduce the costs on the fibercoupling, similar to what had been done in the single chip business. All these efforts are

Multi-Fresnel lenses pumping approach for improving high-power Nd:YAG solar laser beam quality.

Science.gov (United States)

Liang, Dawei; Almeida, Joana

2013-07-20

To significantly improve the present-day high-power solar laser beam quality, a three-stage multi-Fresnel lenses approach is proposed for side-pumping either a Nd:YAG single-crystal or a core-doped Sm(3+)Nd:YAG ceramic rod. Optimum pumping and laser beam parameters are found through ZEMAX and LASCAD numerical analysis. The proposed scheme offers a uniform absorption profile along the rod. 167 W laser power can be achieved, corresponding to 29.3 W/m(2) collection efficiency. High brightness figure of merit of 8.34 W is expected for the core-doped rod within a convex-concave resonator, which is 1300 times higher than that of the most-recent high-power solar laser.

Improvements of the ruby laser oscillator system for laser scattering

International Nuclear Information System (INIS)

Yamauchi, Toshihiko; Kumagai, Katsuaki; Kawakami, Tomohide; Matoba, Tohru; Funahashi, Akimasa

1978-10-01

A ruby laser oscillator system is used to measure electron temperatures of the Tokamak plasmas(JFT-2 and JFT-2a). Improvements have been made of the laser oscillator to obtain the correct values. Described are the improvements and the damages of a ruby rod and a KD*P crystal for Q-switching by laser beam. Improvement are the linear Xe lamp replaced by a helical Xe lamp and in the electrical circuit for Q-switching. The damage of an optical component by a laser beam should be clarified from the damage data; the cause is not found yet. (author)

Morphogenesis of rod-shaped sacculi

NARCIS (Netherlands)

den Blaauwen, T.; de Pedro, M.A.; Nguyen-Distèche, M.; Ayala, J.A.

2008-01-01

For growth and division of rod-shaped bacteria, the cylindrical part of the sacculus has to be elongated and two new cell poles have to be synthesized. The elongation is performed by a protein complex, the elongase that inserts disaccharidepentapeptide units at a limited number of discrete sites

Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

International Nuclear Information System (INIS)

Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

1979-01-01

Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

Control rod

International Nuclear Information System (INIS)

Igarashi, Takao; Yoshimoto, Yuichiro; Sugawara, Satoshi; Fukumoto, Takashi; Endo, Zen-ichiro; Saito, Shozo; Shinpo, Katsutoshi; Nishimura, Akira; Ozawa, Michihiro

1988-01-01

Purpose: To provide a sufficient shutdown margin upon reactor shutdown, prevent sheath deformation without decreasing neutron absorbents and prevent impact shocks exerted to structural materials. Constitution: The control rod of the present invention comprises a neutron absorption region, a sheath deformation means attached to the side wall and means for restricting and supporting axial movement of the neutron absorbent rod. Then, the amount of absorptive nuclei chained absorbents in the lower region is reduced than that in the upper region. In this way, effective neutron absorbing performance can be obtained relative to the neutron importance distribution during reactor shutdown. In addition, since the operationability is improved by reducing the weight of the control rod and the absorptive nuclei chained neutron abosrbers are used, mechanical nuclear life of the control rod can be increased. Thus, it is possible to prevent the outward deformation of the sheath, as well as prevent collision between the neutron absorber rod and the structural material on the side of inserting the control rod generated upon reactor scram by a simple structure. (Kamimura, M.)

Tensile and burst tests in support of the cadmium safety rod failure evaluation

International Nuclear Information System (INIS)

Thomas, J.K.

1992-02-01

The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident

Fuel Rod Vibration Measurement Method using a Flap and its Verification

Energy Technology Data Exchange (ETDEWEB)

Yoo, Joo Young; Park, Nam Gyu; Suh, Jung Min; Jeon, Kyeong Lak [KEPCO NF Co., Daejeon (Korea, Republic of)

2011-10-15

Flow-induced vibration is a critical factor for the mechanical integrity of a fuel rod. This vibration can cause leaked fuel through the mechanism, such as grid to rod fretting. To minimize the failures caused by flow-induced vibration, a robust design is needed which takes into account vibrational characteristics. That is, the spacer grid design should be developed to avoid any excessive vibration. On the one hand, if fuel rod vibration can be measured, an estimation of the excitation forces, which are a critical cause of rod failure, should be possible. Therefore, by applying an external force, flow-induced vibration can be roughly estimated when the fuel rod vibration model is used. KEPCO Nuclear Fuel developed the test loop to research flow-induced vibration as shown in Fig.1. The investigation flow-induced vibration (INFINIT) - the test facility - can measure the grid strap vibration and pressure drop of a 5x5 small scale fuel bundle. Basically, using a Laser Doppler Vibrometer (LDV), the vibration of a structure immersed in high speed fluid can be measured. Grid strap vibration is easily measured using an LDV. However, it is quite difficult to measure fuel rod vibration because of the round surface shape of the rods. In addition, measuring current method using the LDV, it was only possible to directly measure fuel rod vibration at the first row of the bundle as the rods behind the first row are obscured. To solve this problem, a thin flap, as shown in Fig. 2(a) can be used as a reflecting target, gaining access to rods within the bundle. The flap is attached to the fuel rod, as in Fig. 2(b). As a result, most of the inner rod vibration can be measured. Before using a flap to measure fuel rod vibration, a verification process was needed to show whether the LDV signal from the flap vibration provided equivalent and reliable signals. Therefore, impact testing was carried out on the fuel rod using a flap. The LDV signals were then compared with accelerometer

Experimental study of solar pumped laser for magnesium-hydrogen energy cycle

International Nuclear Information System (INIS)

Yabe, T; Okamoto, Y; Ohkubo, T; Uchida, S; Yoshida, K; Bagheri, B; Funatsu, T; Mabuti, A; Oyama, A; Nakagawa, K; Oishi, T; Daito, K; Nakatsuka, M; Yoshida, M; Motokoshi, S; Sato, Y; Baasandash, C; Nakayama, N; Yanagaidani, K

2008-01-01

24.4 W of laser output has been obtained by sun-pumped, Cr-codoped Nd:yttrium aluminum garnet ceramic. The water-cooled laser rod was pumped with a Fresnel lens focusing the natural sunlight. By using the advantages of the Fresnel lenses, the maximum output for unit area of sunlight was 18.7 W/m 2 . Direct concentrated solar illumination was used to pump a 9mm-diameter, 100mm length rod of Cr:Nd:YAG, which was obtained 9%-14% slope efficiency for the laser output. We have analyzed the Cr:YAG laser medium and found it to be an excellent high-power laser candidate for direct solar-pumping schemes which enhances the laser output about 1.8 times more than Nd:YAG

Experimental Breeder Reactor-II automatic control-rod-drive system

International Nuclear Information System (INIS)

Christensen, L.J.

1983-01-01

A computer-controlled automatic control rod drive system (ACRDS) was designed and operated in EBR-II during reactor runs 121 and 122. The ACRDS was operated in a checkout mode during run 121 using a low worth control rod. During run 122 a high worth control rod was used to perform overpower transient tests as part of the LMFBR oxide fuels transient testing program. The testing program required an increase in power of 4 MW/s, a hold time of 12 minutes and a power decrease of 4 MW/s. During run 122, 13 power transients were performed

Control rod drives

International Nuclear Information System (INIS)

Futatsugi, Masao.

1980-01-01

Purpose: To secure the reactor operation safety by the provision of a fluid pressure detecting section for control rod driving fluid and a control rod interlock at the midway of the flow pass for supplying driving fluid to the control rod drives. Constitution: Between a driving line and a direction control valve are provided a pressure detecting portion, an alarm generating device, and a control rod inhibition interlock. The driving fluid from a driving fluid source is discharged by way of a pump and a manual valve into the reactor in which the control rods and reactor fuels are contained. In addition, when the direction control valve is switched and the control rods are inserted and extracted by the control rod drives, the pressure in the driving line is always detected by the pressure detection section, whereby if abnormal pressure is resulted, the alarm generating device is actuated to warn the abnormality and the control rod inhibition interlock is actuated to lock the direction control valve thereby secure the safety operation of the reactor. (Seki, T.)

Control rod drives

International Nuclear Information System (INIS)

Oonuki, Koji.

1981-01-01

Purpose: To increase the driving speed of control rods at rapid insertion with an elongate control rod and an extension pipe while ensuring sufficient buffering performance in a short buffering distance, by providing a plurality of buffers to an extension pipe between a control rod drive source and a control rod in LMFBR type reactor. Constitution: First, second and third buffers are respectively provided to an acceleration piston, an extension pipe and a control rod respectively and the insertion positions for each of the buffers are displaced orderly from above to below. Upon disconnection of energizing current for an electromagnet, the acceleration piston, the extension pipe and the control rod are rapidly inserted in one body. The first, second and third buffers are respectively actuated at each of their falling strokes upon rapid insertion respectively, and the acceleration piston, the extension pipe and the control rod receive the deceleration effect in the order correspondingly. Although the compression force is applied to the control rod only near the stroke end, it does not cause deformation. (Kawakami, Y.)

Calculation of control rods in rectangular reactor, and applications (1960)

International Nuclear Information System (INIS)

Goshen, S.; Pazy, A.

1960-01-01

The aim of this report is to find a method for estimating the anti-reactivity of control rods perpendicular to the axis in a cylindrical pile. The paper is divided into two parts. In the first is given a method of calculating control rods in a rectangular pile, similar to the Nordheim-Scalettar method for cylindrical piles. As an example the formulas are given for the theories of one and two neutron groups, the generalisation for several groups being evident. In the second part we find by a variation method a formula for estimating the Laplacian of a pile, which may be divided into parallelepipeds for which the Laplacian are given. Finally, this formula is used to calculate the anti-reactivity of rods perpendicular to the axis in a cylindrical pile. (author) [fr

Detection and mitigating rod drive control system degradation in Westinghouse PWRs

International Nuclear Information System (INIS)

Gunther, W.; Sullivan, K.

1990-01-01

A study of the effects of aging on the Westinghouse Control Rod Drive (CRD) System was performed as part of the US NRC's Nuclear Plant aging Research (NPAR) Program. For the study, the CRD system boundary includes the power and logic cabinets associated with the manual control rod movement, and the control rod mechanism itself. The aging-related degradation of the interconnecting cables and connectors and the rod position indicating system also were considered. This paper presents the results of that study pertaining to the electrical and instrumentation portions of the CRD system including ways to detect and mitigate system degradation

Magnetostrictive performance of additively manufactured CoFe rods using the LENSTM system

Science.gov (United States)

Jones, Nicholas J.; Yoo, Jin-Hyeong; Ott, Ryan T.; Lambert, Paul K.; Petculescu, Gabriela; Simsek, Emrah; Schlagel, Deborah; Lograsso, Thomas A.

2018-05-01

Magnetostrictive materials exhibit a strain in the presence of a variable magnetic field. While they normally require large, highly oriented crystallographic grains for high strain values, metal additive manufacturing (3D printing) may be able to produce highly textured polycrystalline rods, with properties comparable to those manufactured using the more demanding free standing zone melting (FSZM) technique. Rods of Co75.8Fe24.2 and Co63.7Fe36.3 have been fabricated using the Laser engineered net shaping (LENSTM) system to evaluate the performance of additively manufactured magnetic and magnetostrictive materials. The 76% Co sample showed an average magnetostriction (λ) of 86 ppm at a stress of 124 MPa; in contrast, the 64% Co sample showed only 27 ppm at the same stress. For direct comparison, a Co67Fe33 single crystal disk, also measured as part of this study, exhibited a magnetostriction value of 131 and 91 microstrain in the [100] and [111] directions, respectively, with a calculated polycrystalline value (λs) of 107 microstrain. Electron back scattered diffraction (EBSD) has been used to qualitatively link the performance with crystallographic orientation and phase information, showing only the BCC phase in the 76% Co sample, but three different phases (BCC, FCC, and HCP) in the 64% Co sample.

Magnetostrictive performance of additively manufactured CoFe rods using the LENSTM system

Directory of Open Access Journals (Sweden)

Nicholas J. Jones

2018-05-01

Full Text Available Magnetostrictive materials exhibit a strain in the presence of a variable magnetic field. While they normally require large, highly oriented crystallographic grains for high strain values, metal additive manufacturing (3D printing may be able to produce highly textured polycrystalline rods, with properties comparable to those manufactured using the more demanding free standing zone melting (FSZM technique. Rods of Co75.8Fe24.2 and Co63.7Fe36.3 have been fabricated using the Laser engineered net shaping (LENSTM system to evaluate the performance of additively manufactured magnetic and magnetostrictive materials. The 76% Co sample showed an average magnetostriction (λ of 86 ppm at a stress of 124 MPa; in contrast, the 64% Co sample showed only 27 ppm at the same stress. For direct comparison, a Co67Fe33 single crystal disk, also measured as part of this study, exhibited a magnetostriction value of 131 and 91 microstrain in the [100] and [111] directions, respectively, with a calculated polycrystalline value (λs of 107 microstrain. Electron back scattered diffraction (EBSD has been used to qualitatively link the performance with crystallographic orientation and phase information, showing only the BCC phase in the 76% Co sample, but three different phases (BCC, FCC, and HCP in the 64% Co sample.

Cavitation phenomena in mechanical heart valves: studied by using a physical impinging rod system.

Science.gov (United States)

Lo, Chi-Wen; Chen, Sheng-Fu; Li, Chi-Pei; Lu, Po-Chien

2010-10-01

When studying mechanical heart valve cavitation, a physical model allows direct flow field and pressure measurements that are difficult to perform with actual valves, as well as separate testing of water hammer and squeeze flow effects. Movable rods of 5 and 10 mm diameter impinged same-sized stationary rods to simulate squeeze flow. A 24 mm piston within a tube simulated water hammer. Adding a 5 mm stationary rod within the tube generated both effects simultaneously. Charged-coupled device (CCD) laser displacement sensors, strobe lighting technique, laser Doppler velocimetry (LDV), particle image velocimetry (PIV) and high fidelity piezoelectric pressure transducers measured impact velocities, cavitation images, squeeze flow velocities, vortices, and pressure changes at impact, respectively. The movable rods created cavitation at critical impact velocities of 1.6 and 1.2 m/s; squeeze flow velocities were 2.8 and 4.64 m/s. The isolated water hammer created cavitation at 1.3 m/s piston speed. The combined piston and stationary rod created cavitation at an impact speed of 0.9 m/s and squeeze flow of 3.2 m/s. These results show squeeze flow alone caused cavitation, notably at lower impact velocity as contact area increased. Water hammer alone also caused cavitation with faster displacement. Both effects together were additive. The pressure change at the vortex center was only 150 mmHg, which cannot generate the magnitude of pressure drop required for cavitation bubble formation. Cavitation occurred at 3-5 m/s squeeze flow, significantly different from the 14 m/s derived by Bernoulli's equation; the temporal acceleration of unsteady flow requires further study.

Hydraulic system for the drive of control rod

International Nuclear Information System (INIS)

Niwano, Masao.

1978-01-01

Purpose: To remove thermal stress and improve safety by utilizing water discharged a driving device as a part of cooling water for the device upon driving of control rods. Constitution: A water drain valve is wholly closed and a flow stabilization valve is supplied with an amount of water necessary for driving control rods. Upon driving one control rod, an amount of water required for the driving is caused to flow to the relivant hydraulic control unit and the flow rate in the stabilization valve is reduced by an amount required for the driving to keep the flow rate constant in the flow control valve. Since Excess water conventionally returned to the pressure vessel is utilized as cooling water for the driving device of control rods, the pressure vessel nozzle can be saved. Accordingly, the thermal stress in the nozzle portion can be removed to significantly improve the safety. (Seki, T.)

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 4 discusses the following topics: Rod Compaction/Loading System Test Results and Analysis Report; Waste Collection System Test Results and Analysis Report; Waste Container Transfer Fixture Test Results and Analysis Report; Staging and Cutting Table Test Results and Analysis Report; and Upper Cutting System Test Results and Analysis Report

Study on shadowing effect caused by transient rods at NSRR

International Nuclear Information System (INIS)

Nakamura, T.; Yachi, S.; Ishijima, K.

1992-01-01

Irregularly inserted three control rods created so called shadowing effects on some of the neutronic instruments at the Nuclear Safety Research Reactor (NSRR). During operations at the reactor power of up to 10 MW, the three control rods called transient rods, could be fully or partly inserted into the NSRR core. Reactor power monitors located outside of the core at the direction of deeply inserted transient rods indicated lower power in such operations. Power profiles of the reactor and neutron fluxes at power monitor locations were calculated with a three dimensional neutron diffusion code, CITATION. The calculation indicated that the real reactor power could be smaller than the measured maximum power by as mush as 30 % in such operations. The calculated neutron fluxes well described the changes in the apparent power monitor indications as a function of the transient rod position. (author)

Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5

International Nuclear Information System (INIS)

Pena V, J.D.

2016-01-01

The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

Study on Natural Convection around a vertical heated rod using PIV/LIF technique,

International Nuclear Information System (INIS)

Szijarto, R.; Yamaji, B.; Aszodi, A.

2010-01-01

The Nuclear Training Reactor of the Institute of Nuclear Techniques (Budapest University of Technology and Economics, Hungary) is a pool-type reactor with light water moderator and with a maximum thermal power of 100 kW. The fuel elements are cooled by natural convection. An experimental setup was built to analyse the nature of the natural convection around a heated rod. The flow field was investigated using an electrically heated rod, which models the geometry of a fuel pin in the training reactor. The electric power of the model rod is variable between 0-500 W. The rod was placed in a square-based glass tank. Particle Image Velocimetry and Laser Induced Fluorescence measurement techniques were used to study the velocity and temperature field in a two-dimensional area. The thermal and the hydraulic boundary layers were detected near a rod in a lower section of the aquarium. The laminar-turbulent transition of the flow regime was observed, the maximum velocity of the up-flow was 0.025-0.05 m/s. From the temperature measurements the local heat transfer coefficient was estimated. (Authors)

Stress Analysis of Single Spacer Grid Support considering Fuel Rod

Energy Technology Data Exchange (ETDEWEB)

Yoo, Y. G.; Jung, D. H.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

2010-10-15

Pressurized water reactor (PWR) nuclear fuel assembly is mainly composed of a top-end piece, a bottom-end piece, lots of fuel rods, and several spacer grids. Among them, the main function of spacer grid is protecting fuel rods from Fluid Induced Vibration (FIV). The cross section of spacer grid assembled by laser welding in upper and lower point. When the fuel rod inserted in spacer gird, spring and dimple and around of welded area got a stresses. The main hypothesis of this analysis is the boundary area of HAZ and base metal can get a lot of damage than other area by FIV. So, design factors of spacer grid mainly considered to preventing the fatigue failure in HAZ and spring and dimple of spacer grid. From previous researching, the environment in reactor verified. Pressure and temperature of light water observed 15MPa and 320 .deg. C, and vibration of the fuel rod observed within 0 {approx} 50Hz. In this study, mechanical properties of zirconium alloy that extracted from the test and the spacer grid model which used in the PWR were applied in stress analyzing. General-purpose finite element analysis program was used ANSYS Workbench 12.0.1 version. 3-D CAD program CATIA was used to create spacer grid model

Absorber rod for nuclear reactors in a pebble bed of spherical operating elements

International Nuclear Information System (INIS)

Reinstein, D.; Gnutzmann, H.

1978-01-01

The claim refers to the constructional configuration of an absorber rod, whose and penetrating into the pebble bed has an opening to reduce the fracture rate, so that the operating elements can escape into a channel within the absorber rod. To suit this to the direction of movement of the elements a part of the end of the rod is flexibly connected to the hollow absorber rod via a joint. In this way the mechanical load of the element particles is reduced and simultaneously one achieves that much lower force is required to insert the absorber rod into the pebble bed. (UA) [de

Control rod drive mechanism

International Nuclear Information System (INIS)

Nakamura, Akira.

1981-01-01

Purpose: To ensure the scram operation of a control rod by the reliable detection for the position of control rods. Constitution: A permanent magnet is provided to the lower portion of a connecting rod in engagement with a control rod and a tube having a plurality of lead switches arranged axially therein in a predetermined pitch is disposed outside of the control rod drives. When the control rod moves upwardly in the scram operation, the lead switches are closed successively upon passage of the permanent magnet to operate the electrical circuit provided by way of each of the lead switches. Thus, the position for the control rod during the scram can reliably be determined and the scram characteristic of the control rod can be recognized. (Furukawa, Y.)

Lithium and boron analysis by LA-ICP-MS results from a bowed PWR rod with contact

Directory of Open Access Journals (Sweden)

Puranen Anders

2017-01-01

Full Text Available A previously published investigation of an irradiated fuel rod from the Ringhals 2 PWR, which was bowed to contact with an adjacent rod, identified a significant but highly localised thinning of the clad wall and increased corrosion. Rod fretting was deemed unlikely due to the adhering oxide covering the surfaces. Local overheating in itself was also deemed insufficient to account for the accelerated corrosion. Instead, an enhanced concentration of lithium due to conditions of local boiling was hypothesised to explain the accelerated corrosion. Studsvik has developed a hot cell coupled LA-ICP-MS (Laser Ablation Inductively Coupled Plasma Mass Spectrometer equipment that enables a flexible means of isotopic analysis of irradiated fuel and other highly active surfaces. In this work, the equipment was used to investigate the distribution of lithium (7Li and boron (11B in the outer oxide at the bow contact area. Depth profiling in the clad oxide at the opposite side of the rod to the point of contact, which is considered to have experienced normal operating conditions and which has a typical oxide thickness, evidenced levels of ∼10–20 ppm 7Li and a 11B content reaching hundreds of ppm in the outer parts of the oxide, largely in agreement with the expected range of Li and B clad oxide concentrations from previous studies. In the contact area, the 11B content was similar to the reference condition at the opposite side. The 7Li content in the outermost oxide closest to the contact was, however, found to be strongly elevated, reaching several hundred ppm. The considerable and highly localised increase in lithium content at the area of enhanced corrosion thus offers strong evidence for a case of lithium induced breakaway corrosion during power operation, when rod-to-rod contact and high enough surface heat flux results in a very local increase in lithium concentration.

Modifying a Rodenstock scanning laser ophthalmoscope for imaging densitometry.

Science.gov (United States)

Tornow, R P; Beuel, S; Zrenner, E

1997-08-01

The necessary modifications and technical requirements are described for using a commercially available scanning laser ophthalmoscope (Rodenstock Model 101 SLO) as an imaging densitometer to assess human photopigment distribution. The main requirements are a linear detector amplifier, fast shutters for the laser beams, and a trigger unit. Images must be compensated for varying laser intensity. Both rod and cone photopigments are measured with the 514-nm argon laser of the SLO. Discrimination is possible owing to the different spatial distribution. The cone pigment density peaks in the foveal center (D = 0.40) with a steep decrease with increasing eccentricity E (full width at half-maximum, 2.5 degrees ). Rod photopigment increases with increasing eccentricity (D = 0.23 for E = 11 degrees ). These values are in agreement with previous reported results obtained with scanning laser ophthalmoscopes specially designed for retinal densitometry and high stability.

Apparatus for injection casting metallic nuclear energy fuel rods

Science.gov (United States)

Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

1991-01-01

Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

Method of inserting fuel rod

International Nuclear Information System (INIS)

Kamimoto, Shuji; Imoo, Makoto; Tsuchida, Kenji.

1991-01-01

The present invention concerns a method of inserting a fuel rod upon automatic assembling, automatic dismantling and reassembling of a fuel assembly in a light water moderated reactor, as well as a device and components used therefor. That is, a fuel rod is inserted reliably to an aimed point of insertion by surrounding the periphery of the fuel rod to be inserted with guide rods, and thereby suppressing the movement of the fuel rod during insertion. Alternatively, a fuel rod is inserted reliably to a point of insertion by inserting guide rods at the periphery of the point of insertion for the fuel rod to be inserted thereby surrounding the point of insertion with the guide rods or fuel rods. By utilizing fuel rods already present in the fuel assembly as the guide rods described above, the fuel rod can be inserted reliably to the point of insertion with no additional devices. Dummy fuel rods are previously inserted in a fuel assembly which are then utilized as the above-mentioned guide rods to accurately insert the fuel rod to the point of insertion. (I.S.)

Evaluation of rod insertion issue for NPP Krsko

International Nuclear Information System (INIS)

Gunstek, A.; Kurincic, B.

1998-01-01

The last couple of years incident with control rods sticking in lower part of the fuel assemblies have been reported of several reactor operators and fuel vendors throughout of the world. Several activities were initiated immediately to determine the root cause of incomplete rod insertion. The purpose of this activities were to collect plants trip history data and testing results, review of available worldwide experience, review of plant operation and fuel management, detailed review of manufacturing and material property and to maintain detailed mechanical model. In this paper, we will present activities in Nuclear Power Plant Krsko which have been performed after NRC initiated the Root Cause Process (NRC Bulletin 96-01). NPP Krsko has not experienced rod insertion anomaly yet but anyway the additional tests were carried out. Rod drop time measurements that were performed normally at beginning of cycle at nominal temperature and pressure (HSB mode) have been extended also to end of cycle. Rod drop time, velocity of dropped rods and magnitudes of the initial recoil bounces vs. burnup were also analyzed. Also RCCA drag test with upper internals in place and drive shafts attached to RCCAs has been performed since then. At last two outages (1997 and 1998) drag test were carried out with digital scale meter to gather additional information. In addition to that, the reload core design has been performed with new constrains on rodded fuel assembly burnup as proposed by the industry.(author)

Control rod drives

International Nuclear Information System (INIS)

Hayakawa, Hiroyasu.

1979-01-01

Purpose: To enable rapid control in a simple circuit by providing a motor control device having an electric capacity capable of simultaneously driving all of the control rods rapidly only in the inserting direction as well as a motor controlling device capable of fine control for the insertion and extraction at usual operation. Constitution: The control rod drives comprise a first motor control device capable of finely controlling the control rods both in inserting and extracting directions, a second motor control device capable of rapidly driving the control rods only in the inserting direction, and a first motor switching circuit and a second motor switching circuit switched by switches. Upon issue of a rapid insertion instruction for the control rods, the second motor switching circuit is closed by the switch and the second motor control circuit and driving motors are connected. Thus, each of the control rod driving motors is driven at a high speed in the inserting direction to rapidly insert all of the control rods. (Yoshino, Y.)

Characteristics of a laser beam produced by using thermal lensing effect compensation in a fiber-coupled laser-diode-pumped Nd:YAG ceramic laser

International Nuclear Information System (INIS)

Kim, Duck-Lae; Kim, Byung-Tai

2010-01-01

The characteristics of a laser beam produced by using thermal lensing effect compensation in a fiber-coupled laser-diode Nd:YAG ceramic laser were investigated. The thermal lensing effect was compensated for by using a compensator, which was 25 mm away from the laser rod, with a focal length of 30 mm and an effective clear aperture of 22 mm. Using a compensator, the divergence and the beam propagation factor M 2 of the output beam were 5.5 mrad and 2.4, respectively, under a pump power of 12W. The high-frequency components in the compensated laser beam were removed.

Activation calculation of steel of the control rods of TRIGA Mark III reactor

International Nuclear Information System (INIS)

Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A.

2014-10-01

In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

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.)

Fuel rod simulator effects in flooding experiments single rod tests

International Nuclear Information System (INIS)

Nishida, M.

1984-09-01

The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

Perovskite structures in the formation of nano-rods in REBa2Cu3O7-δ films self-organization to perovskite structures

International Nuclear Information System (INIS)

Mukaida, Masashi; Kai, Hideki; Shingai, Yuki

2009-01-01

Cubic perovskite structure has been found to play an important role for the nano-rod formation in REBa 2 Cu 3 O 7-δ films. BaWO 4 , with a sheelite structure, and BaNb 2 O 6 , with a tungsten bronze structure, were doped into REBa 2 Cu 3 O 7-δ targets. Laser-deposited, these materials form nano-rods in REBa 2 Cu 3 O 7-δ films accompanied by Ln elements, resulting in the composition of a pseudo-cubic perovskite structure. This was confirmed by selected area electron diffraction patterns (SADP) and composition mapping using energy-dispersive X-ray spectroscopy scanning transmission electron microscope (EDS-STEM) analysis. BaWO 4 with a sheelite structure, and BaNb 2 O 6 with a tungsten bronze structure, doped into targets no longer retain their structures, but can form pseudo-cubic perovskite structures in laser-deposited REBa 2 Cu 3 O 7-δ films. The perovskite crystal structure is thought to be important for nano-rod formation in the laser deposited REBa 2 Cu 3 O 7-δ film. (author)

Maximum/minimum asymmetric rod detection

International Nuclear Information System (INIS)

Huston, J.T.

1990-01-01

This patent describes a system for determining the relative position of each control rod within a control rod group in a nuclear reactor. The control rod group having at least three control rods therein. It comprises: means for producing a signal representative of a position of each control rod within the control rod group in the nuclear reactor; means for establishing a signal representative of the highest position of a control rod in the control rod group in the nuclear reactor; means for establishing a signal representative of the lowest position of a control rod in the control rod group in the nuclear reactor; means for determining a difference between the signal representative of the position of the highest control rod and the signal representative of the position of the lowest control rod; means for establishing a predetermined limit for the difference between the signal representative of the position of the highest control rod and the signal representative of the position of the lowest control rod; and means for comparing the difference between the signals with the predetermined limit. The comparing means producing an output signal when the difference between the signals exceeds the predetermined limit

Radial power density distribution of MOX fuel rods in the HBWR

International Nuclear Information System (INIS)

Koo, Yang Hyun; Joo, Hyung Kook; Lee, Byung Ho; Sohn, Dong Seong

1999-07-01

Two MOX fuel rods, which ar being fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with the Korea Atomic Energy Research Institute (KAERI), are going to be irradiated in the HBWR (Halden Boiling Water Reactor) from the beginning of 2000 in the framework of OECD Halden Reactor Programme (HRP) together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is a basic property in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR H BWR that calculates radial power density distribution for three MOX fuel rods have been developed subroutine FACTOR H BWR gives good agreement with the physics calculation except slight underprediction in the central part and a little overprediction at the outer part of the pellet. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. (author). 5 refs., 3 tabs., 24 figs

Control rod drives

International Nuclear Information System (INIS)

Hayakawa, Hiroyasu; Kawamura, Atsuo.

1979-01-01

Purpose: To reduce pellet-clad mechanical interactions, as well as improve the fuel safety. Constitution: In the rod drive of a bwr type reactor, an electric motor operated upon intermittent input such as of pulse signals is connected to a control rod. A resolver for converting the rotational angle of the motor to electric signals is connected to the rotational shaft of the motor and the phase difference between the output signal from the resolver and a reference signal is adapted to detect by a comparator. Based on the detection result, the controller is actuated to control a motor for control rod drive so that fine control for the movement of the control rod is made possible. This can reduce the moving distance of the control rod, decrease the thermal stress applied to the control rod and decrease the pellet clad mechanical interaction failures due to thermal expansion between the cladding tube and the pellets caused by abrupt changes in the generated power. (Furukawa, Y.)

Research on NC laser combined cutting optimization model of sheet metal parts

Science.gov (United States)

Wu, Z. Y.; Zhang, Y. L.; Li, L.; Wu, L. H.; Liu, N. B.

2017-09-01

The optimization problem for NC laser combined cutting of sheet metal parts was taken as the research object in this paper. The problem included two contents: combined packing optimization and combined cutting path optimization. In the problem of combined packing optimization, the method of “genetic algorithm + gravity center NFP + geometric transformation” was used to optimize the packing of sheet metal parts. In the problem of combined cutting path optimization, the mathematical model of cutting path optimization was established based on the parts cutting constraint rules of internal contour priority and cross cutting. The model played an important role in the optimization calculation of NC laser combined cutting.

The nuclear fuel rod character recognition system based on neural network technique

International Nuclear Information System (INIS)

Kim, Woong-Ki; Park, Soon-Yong; Lee, Yong-Bum; Kim, Seung-Ho; Lee, Jong-Min; Chien, Sung-Il.

1994-01-01

The nuclear fuel rods should be discriminated and managed systematically by numeric characters which are printed at the end part of each rod in the process of producing fuel assembly. The characters are used to examine manufacturing process of the fuel rods in the inspection process of irradiated fuel rod. Therefore automatic character recognition is one of the most important technologies to establish automatic manufacturing process of fuel assembly. In the developed character recognition system, mesh feature set extracted from each character written in the fuel rod is employed to train a neural network based on back-propagation algorithm as a classifier for character recognition system. Performance evaluation has been achieved on a test set which is not included in a training character set. (author)

Laser cutting of high manganese cast steel; Komangan chuko no laser setsudan

Energy Technology Data Exchange (ETDEWEB)

Kataoka, Y.; Tokunaga, T. [University of Industrial Technology, Kanagawa (Japan); Miyazaki, T. [Chiba Institute of Technology, Chiba (Japan)

1994-08-25

This paper discusses applicability of CO2 laser to cut high manganese cast steel, and investigates the cutting conditions and characteristics. The tested material is made of steel scrap, ferro Mn and ferro Cr of 50 kg in total, which was deoxidized by using pure aluminum and injected into CO2 die by means of the ordinary casting method to make a circular rod with a diameter of 28 mm. The rod was given a heating and water toughening treatment in a muffle furnace maintaining N2 atmosphere. The base structure is an austenite system of Mn 12.4% by mass with hardness of MHV 220 to 230. The paper describes discussions on test pieces (with a thickness of 2 mm) fabricated under a laser beam frequency of 150 Hz, power outputs of 250, 350 and 500 W, and cutting speeds of 100, 300 and 500 mm/min. The cutting width increases as the laser power is increased, but is not governed by the cutting speed. Increased cutting speed roughens the surface of a cut face. The laser cutting has caused no change in hardness of the base material, and no processing deterioration has been recognized. As described, the laser cutting can be applied to finish-cutting if the cutting condition is selected properly. Simplification of the cutting process and improvement of working environment can be expected from the laser cutting. 14 refs., 12 figs., 2 tabs.

Calculation of drop course of control rod assembly in PWR

International Nuclear Information System (INIS)

Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

2013-01-01

The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

Modelling of Rod No 8 in IFA-597:3

International Nuclear Information System (INIS)

Malen, K.

2002-06-01

A Westinghouse Atom 8x8 fuel rod irradiated in the Ringhals 1 BWR for 12 years to a local burnup of about 67 MWd/kgU was refabricated, instrumented with centreline thermocouple and pressure transducer, and irradiated in IFA-597.2 for about 20 days and in IFA-597.3 for about four months. The rod was then sent to Kjeller for puncturing and then to the Studsvik hot cells for detailed post-irradiation examinations. The peak centreline, temperature was close to 1350 deg C. The total fission gas release (FGR) determined from the puncturing was approximately 20 %. Electron probe microanalysis on a fuel section from the central part of the rod showed that virtually 100 % Xe release had occurred in the central part of the pellet out to about half the pellet radius, and this thermal release from the central part of the fuel accounted for the measured total FGR. Optical and scanning electron microscopy of the fuel cross-section showed complete pellet-clad bonding as well as an extensive high burnup 'rim' structure extending at least 0,15 mm in from the fuel surface. The fuel microstructure was characterised at different radial positions in the pellet. This report describes modelling of the rod behaviour using the code SKIROD, in particular fuel temperature and fission gas release. The transient response of the fuel centre line temperature after a scram is also modelled using the code TOODEE2. The modelling results are compared to the experimental results

Motion simulation of hydraulic driven safety rod using FSI method

International Nuclear Information System (INIS)

Jung, Jaeho; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In

2013-01-01

Hydraulic driven safety rod which is one of them is being developed by Division for Reactor Mechanical Engineering, KAERI. In this paper the motion of this rod is simulated by fluid structure interaction (FSI) method before manufacturing for design verification and pump sizing. A newly designed hydraulic driven safety rod which is one of reactivity control mechanism is simulated using FSI method for design verification and pump sizing. The simulation is done in CFD domain with UDF. The pressure drop is changed slightly by flow rates. It means that the pressure drop is mainly determined by weight of moving part. The simulated velocity of piston is linearly proportional to flow rates so the pump can be sized easily according to the rising and drop time requirement of the safety rod using the simulation results

Thermal hydraulic performance of naturally aspirated control rod housing assemblies

International Nuclear Information System (INIS)

Geiger, G.T.; Randolph, H.W.; Paik, I.K.; Foti, D.J.

1992-01-01

Savannah River Site reactors are comprised of heat generating fuel/target assemblies, control rods which regulate reactor power, and heavy water which acts as the coolant and as a moderator. The fuel/target assemblies are cooled by the downflow of heavy water while the control rods are cooled via upflow. Five control rods are grouped with two safety rods in seven-channel assemblies called septifoils. Under normal operating conditions, the reactor power level, radial shape flux and axial power flux are regulated by the positioning of the control rods. The control rods are solid rods of a lithium-aluminum alloy with an thin aluminum outer sheath. Lithium is a good absorber of neutrons and, thus control rod temperatures rise with reactor power. At conditions of sufficiently high reactor power and degraded coolant flow, the control rods could heat sufficiently to cause a metallurigical failure of the sheath leading to molten material coming in contact with water and the possibility of a steam explosion. An accident has been postulated as part of the analysis involving the safety upgrade of Savannah River Site reactors in which the housing is not seated on the pin. Coolant from the upflow pin would not be directed into the housing but, into the moderator space surrounding the housing. Only naturally aspirated cooling due to buoyancy effects would be available to cool the control rods and the coolant mass flow rate would drop significantly from its nominal value. In this study, the mechanisms and limits of cooling heated rods housed in an unseated septifoil are addressed. Experiments were conducted on a shortened, prototypic housing with electrically heated rods to gain an understanding of the phenomena governing the cooling in such a case and develop data which can be used to evaluate predictive models. These experiments are described, their results discussed, and the predictions of current models is presented

Anomaly Detection In Additively Manufactured Parts Using Laser Doppler Vibrometery

Energy Technology Data Exchange (ETDEWEB)

Hernandez, Carlos A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-09-29

Additively manufactured parts are susceptible to non-uniform structure caused by the unique manufacturing process. This can lead to structural weakness or catastrophic failure. Using laser Doppler vibrometry and frequency response analysis, non-contact detection of anomalies in additively manufactured parts may be possible. Preliminary tests show promise for small scale detection, but more future work is necessary.

Status of rod consolidation

International Nuclear Information System (INIS)

Bailey, W.J.

1985-04-01

Two of the factors that need to be taken into account with rod consolidation are (1) the effects on rods from their removal from the fuel assembly and (2) the effects on rods as a result of the consolidation process. Potential components of both factors are described in the report. Discussed under (1) are scratches on the fuel rod surfaces, rod breakage, crud, extended burnup, and possible cladding embrittlement due to hydrogen injection at BWRs. Discussed under (2) are the increased water temperature (less than 10 0 C) because of closer packing of the rods, formation of crevices between rods in the close-packed mode, contact with dissimilar metals, and the potential for rapid heating of fuel rods following the loss of water from a spent fuel storage pool. Another factor that plays an important role in rod consolidation is the cost of disposal of the nonfuel-bearing components of the fuel assembly. Also, the dose rate from the components - especially Inconel spacer grids - can affect the handling procedures. Several licensing issues that exist are described. A list of recommendations is provided. 98 refs., 5 figs., 5 tabs

TECHNICAL NOTE: Characteristic analysis of an ultrasonic micromotor using a 3 mm diameter piezoelectric rod

Science.gov (United States)

Chu, Xiangcheng; Yan, Li; Li, Longtu

2004-04-01

Smart systems and devices generally use certain microstructures, e.g. rod- and strip-shaped structures. In this paper, a miniaturized piezoelectric rod is analysed using the finite element method (FEM) and a laser scanning vibrometer (LSV). The effects of some factors, including the detailed structure, material parameters and input voltage, on the resonant frequencies and vibration behaviors of a piezoelectric rod are studied. On the basis of experimental results, the vibration modes of the piezoelectric rod can be made available for use in fabricating an ultrasonic micromotor or piezoelectric actuators of other types. The prototype motor fabricated here has a maximum output torque of 410 µN m for a stainless steel stator and 360 µN m for a copper stator. This article was originally published in 2003 by the Israel Academy of Sciences and Humanities in the framework of its Albert Einstein Memorial Lectures series. Reprinted by permission of the Israel Academy of Sciences and Humanities.

Laser Welding Test Results with Gas Atmospheres in Welding Chamber

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang-Young; Hong, Jin-Tae; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seo-Yun; Yang, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The weld beads of specimens welded under identical conditions in the helium and argon gas were cleaner, more regular, and steadier than those in a vacuum. The penetration depth of the FZ in the vacuum was much deeper than those in the helium and argon gas. To measure the irradiation properties of nuclear fuel in a test reactor, a nuclear fuel test rod instrumented with various sensors must be fabricated with assembly processes. A laser welding system to assemble the nuclear fuel test rod was designed and fabricated to develop various welding technologies of the fuel test rods to joint between a cladding tube and end-caps. It is an air-cooling optical fiber type and its emission modes are a continuous (CW) mode of which the laser generates continuous emission, and pulse (QCW) mode in which the laser internally generates sequences of pulses. We considered the system welding a sample in a chamber that can weld a specimen in a vacuum and inert gas atmosphere, and the chamber was installed on the working plate of the laser welding system. In the chamber, the laser welding process should be conducted to have no defects on the sealing area between a cladding tube and an end-cap.

Control rod drive

International Nuclear Information System (INIS)

Okutani, Tetsuro.

1988-01-01

Purpose: To provide a simple and economical control rod drive using a control circuit requiring no pulse circuit. Constitution: Control rods in a BWR type reactor are driven by hydraulic pressure and inserted or withdrawn in the direction of applying the hydraulic pressure. The direction of the hydraulic pressure is controlled by a direction control valve. Since the driving for the control rod is extremely important in view of the operation, a self diagnosis function is disposed for rapid inspection of possible abnormality. In the present invention, two driving contacts are disposed each by one between the both ends of a solenoid valve of the direction control valve for driving the control rod and the driving power source, and diagnosis is conducted by alternately operating them. Therefore, since it is only necessary that the control circuit issues a driving instruction only to one of the two driving contacts, the pulse circuit is no more required. Further, since the control rod driving is conducted upon alignment of the two driving instructions, the reliability of the control rod drive can be improved. (Horiuchi, T.)

Control rod withdrawal monitoring device

International Nuclear Information System (INIS)

Ebisuya, Mitsuo.

1984-01-01

Purpose: To prevent the power ramp even if a plurality of control rods are subjected to withdrawal operation at a time, by reducing the reactivity applied to the reactor. Constitution: The control rod withdrawal monitoring device is adapted to monitor and control the withdrawal of the control rods depending on the reactor power and the monitoring region thereof is divided into a control rod group monitoring region a transition region and a control group monitoring not interfere region. In a case if the distance between a plurality of control rods for which the withdrawal positions are selected is less than a limiting value, the coordinate for the control rods, distance between the control rods and that the control rod distance is shorter are displayed on a display panel, and the withdrawal for the control rods are blocked. Accordingly, even if a plurality of control rods are subjected successively to the withdrawal operation contrary to the control rod withdrawal sequence upon high power operation of the reactor, the power ramp can be prevented. (Kawakami, Y.)

Releasing method of connection of control rod and its drive mechanism in a reactor

International Nuclear Information System (INIS)

Ishida, Kazuo; Futatsugi, Masao.

1976-01-01

Object: To disengage a control rod from a control rod drive device in a boiling water reactor with a minimal failure of the device, when connection there between cannot be released in a normal manner. Structure: First, a part of a piston tube in the control rod drive device is withdrawn externally of a control rod housing and cut. Next, a discharge tool, which is designed to be connected with the cut piston tube, is connected to the remainder of the piston tube within the housing and the aforesaid piston tube is pushed into the index tube. The index tube is then cut by the discharge tool. Thus, the control rod drive device and the control rod may be separated. Thereafter, the control rod may be removed from the top of the reactor container whereas the control rod drive device removed from the bottom thereof. (Ikeda, J.)

Wave Stresses in the Anvil Hammer Rods under Impact Including Ram Mass and Deformation Force of Forgings

Directory of Open Access Journals (Sweden)

V. M. Sinitskiy

2016-01-01

Full Text Available When operating the anvil hammers there occur impacts of die tooling and as a consequence, virtually instantaneous impact stops of motion of drop hammer parts. Such operating conditions come with accelerated failures of the anvil hammer rods because of emerging significant wave stresses. Engineering practice widely uses variation, difference, and integral methods to calculate wave stresses. However, to use them a researcher has to acquire certain skills, and the special programs should be available. The paper considers a method for estimating the wave stress changes in the anvil hammer rods, which is based on the wave equation of the Laplace transform. It presents a procedure for generating differential equations and their solution using the operator method. These equations describe the wave processes of strain and stress propagation in the anvil hammer rod under non-rigid impact with the compliance obstacle of the drop hammer parts. The work defines how the piston and rod mass and also the mechanical and geometric parameters of the rod influence on the stress level in the rod sealing of the hammer ram. Analysis of the results shows that the stresses in the rod sealing are proportional to the total amount of wave stresses caused by the rod and piston impact included in the total weight of the system. The piston influence on the stresses in the rod under impact is in direct proportion to the ratio of its mass to the mass of the rod. Geometric parameters of the rod and speed of drop parts before the impact influence on the stress value as well. It was found that if the time of impact is less than the time of the shock wave running in forward and backward direction, the impact with a compliance obstacle is equivalent to that of with a rigid obstacle, and the dependence of the wave stresses follows the Zhukovsky formula of direct pressure shock. The presented method of stress calculation can be successfully used to select the optimal mass and the rod

Investigation of axial power gradients near a control rod tip

Energy Technology Data Exchange (ETDEWEB)

Loberg, John, E-mail: John.Loberg@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Osterlund, Michael, E-mail: Michael.Osterlund@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Bejmer, Klaes-Hakan, E-mail: Klaes-Hakan.Bejmer@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Blomgren, Jan, E-mail: Jan.Blomgren@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Kierkegaard, Jesper, E-mail: Jesper.Kierkegaar@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden)

2011-07-15

Highlights: > Pin power gradients near BWR control rod tips have been investigated. > A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. > Small nodes increases pin power gradients; standard nodes underestimates gradients. > The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, {approx}15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

Investigation of axial power gradients near a control rod tip

International Nuclear Information System (INIS)

Loberg, John; Osterlund, Michael; Bejmer, Klaes-Hakan; Blomgren, Jan; Kierkegaard, Jesper

2011-01-01

Highlights: → Pin power gradients near BWR control rod tips have been investigated. → A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. → Small nodes increases pin power gradients; standard nodes underestimates gradients. → The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, ∼15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

Prototypical spent nuclear fuel rod consolidation equipment: Phase 2, Final design report: Volume 4, Appendices: Part 3

International Nuclear Information System (INIS)

Ciez, A.P.

1987-01-01

The purpose of this manual is to provide assembly, installation, operation, maintenance, and off-normal recovery procedures for the Consolidation Equipment. The Consolidation System is a horizontal, dry system capable of processing one Pressurized Water Reactor (PWR) fuel assembly or one Boiling Water Reactor (BWR) fuel assembly at a time. The system will process all spent PWR and BWR fuels from the commercial US nuclear power reactor industry. Component changeouts for various fuel types have been minimized to reduce costs, required in-cell module storage space, and to increase efficiency by decreasing set-up time between fuel consolidation campaigns. The most important feature of the Westinghouse system is the ability to control the fuel rods at all times during the consolidation process from rod extraction, through canister loading. This features assures that the rods from two PWR fuel assemblies or four BWR fuel assemblies (minimum) can be loaded into one consolidated rods canister

Stress analysis and optimization of Nd:YAG pulsed laser processing of notches for fracture splitting of a C70S6 connecting rod

International Nuclear Information System (INIS)

Kou, Shuqing; Gao, Yan; Zhao, Yong; Lin, Baojun

2017-01-01

The pulsed laser pre-processing of a notch as the fracture initiation source for the splitting process is the key mechanism of an advanced fracture splitting technology for C70S6 connecting rods. This study investigated the stress field of Nd:YAG pulsed laser grooving, which affects the rapid fracture initiation at the notch root and the controlled crack extension in the critical fracture splitting quality, to improve manufacturing quality. Thermal elastic-plastic incremental theory was applied to build the finite element analysis model of the stress field of pulsed laser grooving for fracture splitting based on the Rotary-Gauss body heat source. The corresponding numerical simulation of the stress field was conducted. The changes and distributions of the stress during pulsed laser grooving were examined, the influence rule of the primary technological parameters on the residual stress was analyzed, and the analysis results were validated by the corresponding cutting experiment. Results showed that the residual stress distribution was concentrated in the Heat-affected zone (HAZ) near the fracture splitting notch, which would cause micro-cracks in the HAZ. The stress state of the notch root in the fracture initiation direction was tensile stress, which was beneficial to the fracture initiation and the crack rapid extension in the subsequent fracture splitting process. However, the uneven distribution of the stress could lead to fracture splitting defects, and thus the residual stress should be lowered to a reasonable range. Decreasing the laser pulse power, increasing the processing speed, and lowering the pulse width can lower the residual stress. Along with the actual production, the reasonable main technological parameters were obtained.

Stress analysis and optimization of Nd:YAG pulsed laser processing of notches for fracture splitting of a C70S6 connecting rod

Energy Technology Data Exchange (ETDEWEB)

Kou, Shuqing; Gao, Yan; Zhao, Yong; Lin, Baojun [Jilin University, Changchun (China)

2017-05-15

The pulsed laser pre-processing of a notch as the fracture initiation source for the splitting process is the key mechanism of an advanced fracture splitting technology for C70S6 connecting rods. This study investigated the stress field of Nd:YAG pulsed laser grooving, which affects the rapid fracture initiation at the notch root and the controlled crack extension in the critical fracture splitting quality, to improve manufacturing quality. Thermal elastic-plastic incremental theory was applied to build the finite element analysis model of the stress field of pulsed laser grooving for fracture splitting based on the Rotary-Gauss body heat source. The corresponding numerical simulation of the stress field was conducted. The changes and distributions of the stress during pulsed laser grooving were examined, the influence rule of the primary technological parameters on the residual stress was analyzed, and the analysis results were validated by the corresponding cutting experiment. Results showed that the residual stress distribution was concentrated in the Heat-affected zone (HAZ) near the fracture splitting notch, which would cause micro-cracks in the HAZ. The stress state of the notch root in the fracture initiation direction was tensile stress, which was beneficial to the fracture initiation and the crack rapid extension in the subsequent fracture splitting process. However, the uneven distribution of the stress could lead to fracture splitting defects, and thus the residual stress should be lowered to a reasonable range. Decreasing the laser pulse power, increasing the processing speed, and lowering the pulse width can lower the residual stress. Along with the actual production, the reasonable main technological parameters were obtained.

Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

International Nuclear Information System (INIS)

Lima, R. de C.F. de.

1984-01-01

In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author) [pt

A 40 W cw Nd:YAG solar laser pumped through a heliostat: a parabolic mirror system

International Nuclear Information System (INIS)

Almeida, J; Liang, D; Guillot, E; Abdel-Hadi, Y

2013-01-01

Solar-pumped solid-state lasers are promising for renewable extreme-temperature material processing. Here, we report a significant improvement in solar laser collection efficiency by pumping the most widely used Nd:YAG single-crystal rod through a heliostat–parabolic mirror system. A conical-shaped fused silica light guide with 3D-CPC output end is used to both transmit and compress the concentrated solar radiation from the focal zone of a 2 m diameter parabolic mirror to a 5 mm diameter Nd:YAG rod within a conical pump cavity, which enables multi-pass pumping through the laser rod. 40 W cw laser power is measured, corresponding to 13.9 W m −2 record-high collection efficiency for the solar laser pumped through a heliostat–parabolic mirror system. 2.9% slope efficiency is fitted, corresponding to 132% enhancement over that of our previous pumping scheme. A 209% reduction in threshold pump power is also registered. (paper)

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 8 discusses Control System SOT Tests Results and Analysis Report. This is a continuation of Book 7

Examination of cadmium safety rod thermal test specimens and failure mechanism evaluation

International Nuclear Information System (INIS)

Thomas, J.K.; Peacock, H.B.; Iyer, N.C.

1992-01-01

The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of a hypothetical LOCA event. Accordingly, an experimental cadmium safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. Companion reports describe the experiments and a structural evaluation (finite element analysis) of the safety rod. This report deals primarily with the examination of the test specimens, evaluation of possible failure mechanisms, and confirmatory separate effects experiments. It is concluded that the failures observed in the cadmium safety rod thermal tests which occurred at low temperature (T 800 degrees C) with fast thermal ramp rates are concluded to be mechanical in nature without significant environmental degradation. Based on these tests, tasks were initiated to design and manufacture B 4 C safety rods to replace the cadmium safety rods. The B 4 C safety rods have been manufactured at this time and it is currently planned to charge them to the reactor in the near future. 60 refs

Knowledge based system for control rod programming of BWRs

International Nuclear Information System (INIS)

Fukuzaki, Takaharu; Yoshida, Ken-ichi; Kobayashi, Yasuhiro

1988-01-01

A knowledge based system has been developed to support designers in control rod programming of BWRs. The programming searches through optimal control rod patterns to realize safe and effective burning of nuclear fuel. Knowledge of experienced designers plays the main role in minimizing the number of calculations by the core performance evaluation code. This code predicts power distibution and thermal margins of the nuclear fuel. This knowledge is transformed into 'if-then' type rules and subroutines, and is stored in a knowledge base of the knowledge based system. The system consists of working area, an inference engine and the knowledge base. The inference engine can detect those data which have to be regenerated, call those subroutine which control the user's interface and numerical computations, and store competitive sets of data in different parts of the working area. Using this system, control rod programming of a BWR plant was traced with about 500 rules and 150 subroutines. Both the generation of control rod patterns for the first calculation of the code and the modification of a control rod pattern to reflect the calculation were completed more effectively than in a conventional method. (author)

Preliminary design and manufacturing feasibility study for a machined Zircaloy triangular pitch fuel rod support system (grids) (AWBA development program)

International Nuclear Information System (INIS)

Horwood, W.A.

1981-07-01

General design features and manufacturing operations for a high precision machined Zircaloy fuel rod support grid intended for use in advanced light water prebreeder or breeder reactor designs are described. The grid system consists of a Zircaloy main body with fuel rod and guide tube cells machined using wire EDM, a separate AM-350 stainless steel insert spring which fits into a full length T-slot in each fuel rod cell, and a thin (0.025'' or 0.040'' thick) wire EDM machined Zircaloy coverplate laser welded to each side of the grid body to retain the insert springs. The fuel rods are placed in a triangular pitch array with a tight rod-to-rod spacing of 0.063 inch nominal. Two dimples are positioned at the mid-thickness of the grid (single level) with a 90 0 included angle. Data is provided on the effectiveness of the manufacturing operations chosen for grid machining and assembly

Nuclear fuel rod loading apparatus

International Nuclear Information System (INIS)

King, H.B.; Macivergan, R.; Mckenzie, G.W.

1980-01-01

An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

Control rod shutdown system

International Nuclear Information System (INIS)

Miyamoto, Yoshiyuki; Higashigawa, Yuichi.

1996-01-01

The present invention provides a control rod terminating system in a BWR type nuclear power plant, which stops an induction electric motor as rapidly as possible to terminate the control rods. Namely, the control rod stopping system controls reactor power by inserting/withdrawing control rods into a reactor by driving them by the induction electric motor. The system is provided with a control device for controlling the control rods and a control device for controlling the braking device. The control device outputs a braking operation signal for actuating the braking device during operation of the control rods to stop the operation of the control rods. Further, the braking device has at least two kinds of breaks, namely, a first and a second brakes. The two kinds of brakes are actuated by receiving the brake operation signals at different timings. The brake device is used also for keeping the control rods after the stopping. Even if a stopping torque of each of the breaks is small, different two kinds of brakes are operated at different timings thereby capable of obtaining a large stopping torque as a total. (I.S.)

DESIGN OF WIRE-WRAPPED ROD BUNDLE MATCHED INDEX-OF-REFRACTION EXPERIMENTS

Energy Technology Data Exchange (ETDEWEB)

Hugh McIlroy; Hongbin Zhang; Kurt Hamman

2008-05-01

Experiments will be conducted in the Idaho National Laboratory (INL) Matched Index-of-Refraction (MIR) Flow Facility [1] to characterize the three-dimensional velocity and turbulence fields in a wire-wrapped rod bundle typically employed in liquid-metal cooled fast reactors and to provide benchmark data for computer code validation. Sodium cooled fast reactors are under consideration for use in the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program. The experiment model will be constructed of quartz components and the working fluid will be mineral oil. Accurate temperature control (to within 0.05 oC) matches the index-of-refraction of mineral oil with that of quartz and renders the model transparent to the wavelength of laser light employed for optical measurements. The model will be a scaled 7-pin rod bundle enclosed in a hexagonal canister. Flow field measurements will be obtained with a LaVision 3-D particle image velocimeter (PIV) and complimented by near-wall velocity measurements obtained from a 2-D laser Doppler velocimeter (LDV). These measurements will be used as benchmark data for computational fluid dynamics (CFD) validation. The rod bundle model dimensions will be scaled up from the typical dimensions of a fast reactor fuel assembly to provide the maximum Reynolds number achievable in the MIR flow loop. A range of flows from laminar to fully-turbulent will be available with a maximum Reynolds number, based on bundle hydraulic diameter, of approximately 22,000. The fuel pins will be simulated by 85 mm diameter quartz tubes (closed on the inlet ends) and the wire-wrap will be simulated by 25 mm diameter quartz rods. The canister walls will be constructed from quartz plates. The model will be approximately 2.13 m in length. Bundle pressure losses will also be measured and the data recorded for code comparisons. The experiment design and preliminary CFD calculations, which will be used to provide qualitative hydrodynamic

Comparison with experiment of COMETHE III-L fuel rod behaviour predictions

International Nuclear Information System (INIS)

Vliet, J. van; Billaux, M.

1983-01-01

A comparison is presented between experimental results and COMETHE III-L fuel rod behaviour predictions. The first part of the paper focuses on mechanical aspects, with as main experiments, AECL X-264 and Studsvik Interramp. The second part presents the results of a wide FGR benchmarking campaign, with a reference to previous COMETHE versions. It appears that the variance between experiment and calculation has decreased by a factor four when the III-J version was improved into the III-L version. As conclusion, some COMETHE III-L calculations are presented in order to illustrate its capability of predicting fuel rod performance limits. (author)

Fuel rod leak detector

International Nuclear Information System (INIS)

Womack, R.E.

1978-01-01

A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

Development of non-destructive examination system for irradiated fuel rods

International Nuclear Information System (INIS)

Sumerling, R.; Goldsmith, L.A.; Cross, M.T.; McKee, F.

1978-12-01

The development of non-destructive examination (NDE) system for irradiated fuel rods is described. The system is used for testing rods within a concrete cave and consists of three parts: a fully-automated fuel rod-drive machine, designed for easy maintenance; a series of plug-in NDE modules which fit into the central space provided in the machine, plus optical/TV viewing devices and gamma-scan equipment lined up on the rod; and on electronic control equipment situated outside the concrete shielding. The equipment is at present routinely used for viewing, eddy-current testing, gamma-scanning and diameter measurement of rods. The system is flexible in that additional modules can be added later as they are developed, since there is room for three modules of standard size (about 10cm x 10 cm x 3cm) in the machine or one large module taking the full space. New developments include the use of dual frequency eddy-current testing, which allows much greater discrimination against unwanted signals, and measurement of oxide thickness using a high frequency eddy-current probe. (author)

Fuel rods

International Nuclear Information System (INIS)

Hattori, Shinji; Kajiwara, Koichi.

1980-01-01

Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

Control-rod interference effects observed during reactor physics experiments with nuclear ship 'MUTSU'

International Nuclear Information System (INIS)

Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro.

1993-01-01

The control rods in the reactor of the nuclear ship MUTSU are classified into four groups: groups G1 and G2 are located in the central part of the core, while groups G3 and G4 are in the peripheral zone of the core. Several types of mutual interference effects among these control-rod groups were observed during reactor physics experiments with this reactor. During normal hot operations, positive shadowing was dominant between the G1 and G2 groups; the degree of the shadowing effect of one rod group depended on the position of the other rod group. Both positive and negative shadowing effects occurred between an inner rod group (G1 or G2) and an outer group (G3 or G4) depending on the three-dimensional arrangement of the control rods. The rod worths of G1 and G2 increased as a result of slight core burnup, about 1,400 MWd/t, mainly due to the decrease in shadowing effects resulting from a change in control-rod pattern. A three-dimensional diffusion calculation with internal control-rod boundary conditions has proved to be useful for analyzing these various interaction effects. (author)

A study of the deformability of composite rods with end fittings for truss structures

Energy Technology Data Exchange (ETDEWEB)

Stepanychev, E I; Novikov, V V; Sukhanov, A V; Lapotkin, V A; Postnov, A N [Moskovskii Aviatsionnyi Tekhnologicheskii Institut, Moscow (USSR)

1989-04-01

A method is proposed for studying the deformability characteristics of composite rods with end fittings under static loading. Three types of connectors for tubular composite rods are examined from the standpoint of deformability, and differences in the deformability of the central part of a rod and of the connection zone are discussed. The effect of thermal cycling on the deformability of structures of this type is analyzed. 7 refs.

Reduction of Surface Roughness by Means of Laser Processing over Additive Manufacturing Metal Parts.

Science.gov (United States)

Alfieri, Vittorio; Argenio, Paolo; Caiazzo, Fabrizia; Sergi, Vincenzo

2016-12-31

Optimization of processing parameters and exposure strategies is usually performed in additive manufacturing to set up the process; nevertheless, standards for roughness may not be evenly matched on a single complex part, since surface features depend on the building direction of the part. This paper aims to evaluate post processing treating via laser surface modification by means of scanning optics and beam wobbling to process metal parts resulting from selective laser melting of stainless steel in order to improve surface topography. The results are discussed in terms of roughness, geometry of the fusion zone in the cross-section, microstructural modification, and microhardness so as to assess the effects of laser post processing. The benefits of beam wobbling over linear scanning processing are shown, as heat effects in the base metal are proven to be lower.

Behavior of a nine-rod PWR bundle under power-cooling-mismatch conditions

International Nuclear Information System (INIS)

Gunnerson, F.S.; Sparks, D.T.

1979-01-01

An experiment to characterize the behavior of a nine-rod pressurized water reactor (PWR) fuel bundle operating during power-cooling-mismatch (PCM) conditions has been conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The experiment, designated Test PCM-5, is part of a series of PCM experiments designed to evaluate light water reactor (LWR) fuel rod response under postulated accident conditions. Test PCM-5 was the first nine-rod bundle experiment in the PCM test series. The primary objectives and the results of the experiment are described

Construction and testing of a 'fork' type central rod

International Nuclear Information System (INIS)

Guicherd, R.; Meunier, C.

1964-01-01

Aim of the control rod: - Reduction of the flux peaks - Increase of the efficiency - simplification of the mechanics - without modification of the grid nor of the standard fuel plates. Results of the tests: - Reduction of the flux peaks by 29 per cent - Increase of the efficiency by 20 per cent - The suppleness of the plates and the decrease in the number of parts results in greater safety during operation. Conclusions: Study of a generalized adaptation of this type of rod which presents definite advantages. (authors) [fr

Fiber Lasers V

DEFF Research Database (Denmark)

Hansen, Kim P.; Olausson, Christina Bjarnal Thulin; Broeng, Jes

2008-01-01

laser system. We present the latest advancements within airclad fiber technology including a new 70 μm single-mode polarization-maintaining rod-type fiber capable of amplifying to MW power levels. Furthermore we describe the novel airclad based pump combiners and their use in a completely monolithic 350...

Tokay gecko photoreceptors achieve rod-like physiology with cone-like proteins.

Science.gov (United States)

Zhang, Xue; Wensel, Theodore G; Yuan, Ching

2006-01-01

The retinal photoreceptors of the nocturnal Tokay gecko (Gekko gekko) consist exclusively of rods by the criteria of morphology and key features of their light responses. Unlike cones, they display robust photoresponses and have relatively slow recovery times. Nonetheless, the major and minor visual pigments identified in gecko rods are of the cone type by sequence and spectroscopic behavior. In the ongoing search for the molecular bases for the physiological differences between cones and rods, we have characterized the molecular biology and biochemistry of the gecko rod phototransduction cascade. We have cloned cDNAs encoding all or part of major protein components of the phototransduction cascade by RT-PCR with degenerate oligonucleotides designed to amplify cone- or rod-like sequences. For all proteins examined we obtained only cone-like and never rod-like sequences. The proteins identified include transducin alpha (Galphat), phosphodiesterase (PDE6) catalytic and inhibitory subunits, cyclic nucleotide-gated channel (CNGalpha) and arrestin. We also cloned cDNA encoding gecko RGS9-1 (Regulator of G Protein Signaling 9, splice variant 1), which is expressed in both rods and cones of all species studied but is typically found at 10-fold higher concentrations in cones, and found that gecko rods contain slightly lower RGS9-1 levels than mammalian rods. Furthermore, we found that the levels of GTPase accelerating protein (GAP) activity and cyclic GMP (cGMP) phosphodiesterase activity were similar in gecko and mammalian rods. These results place substantial constraints on the critical changes needed to convert a cone into a rod in the course of evolution: The many features of phototransduction molecules conserved between those expressed in gecko rods and those expressed in cones cannot explain the physiological differences, whereas the higher levels of RGS9-1 and GAP activity in cones are likely among the essential requirements for the rapid photoresponses of cones.

Multiple fuel rod gripper

International Nuclear Information System (INIS)

Shields, E.P.

1987-01-01

An apparatus is described for gripping an array of rods comprising: (a) gripping members grippingly engageable with the rods, each of which has a hollow portion terminating in an open end for receiving the end of one of the rods; (b) a closing means for causing the hollow portion of each of the gripping members to apply substantially the same gripping force onto the end of its respective rod, including (i) a locking plate having a plurality of tapered holes registrable with the array of rods, wherein the exterior of each of the gripping members is tapered and nested within one of the tapered holes, (ii) a withdrawing means having a hydraulic plunger operatively connected to each of the gripping members for applying a substantially identical withdrawing force on each of the gripping members, whereby the hollow portion of each of the gripping members applies substantially the same gripping force on its respective rod, and (c) means for detecting whether each of the gripping members has grippingly engaged its respective rod

Development of longer Nd:LGGG crystal for high power laser application

Science.gov (United States)

Yin, Yanru; Tian, Hanlin; Zhang, Jian; Mu, Wenxiang; Zhang, Baitao; Jia, Zhitai; He, Jingliang; Tao, Xutang

2017-11-01

In order to further improve the Nd3+:(LuxGd1-x)3Ga5O12 (Nd:LGGG) crystal performance in high power laser field, a long Nd:LGGG crystal with dimensions of Φ 23 × 112 mm3 has been grown successfully by the Czochralski (Cz) method for laser rod fabrication. Compared with the normal size LGGG crystals (like 30-50 mm in length), we overcame several difficulties in the growth of longer ones, including crystal cracking by a large longitudinal temperature gradient, spiral growth by a small radial temperature gradient, and growth instability and even constitutional super cooling by Ga2O3 volatilizing continuously. The doping concentrations of Nd3+ and Lu3+ in the as-grown crystal and the crystal optical quality have been measured. The performance of diode-side-pumped Nd:LGGG rod laser has been preliminarily tested for the first time, simply by replacing the Nd:YAG crystal rod inside a commercial laser module. Under an incident pump power of 160 W, the maximum continuous wave output power of 38 W has been obtained, corresponding to an optical-optical conversion efficiency of 23.8% and a slope efficiency of 40.8%, respectively.

High-yield production of hydrophobins RodA and RodB from Aspergillus fumigatus in Pichia pastoris

DEFF Research Database (Denmark)

Pedersen, Mona Højgaard; Borodina, Irina; Moresco, Jacob Lange

2011-01-01

A as well as rRodB were able to convert a glass surface from hydrophilic to hydrophobic similar to native RodA, but only rRodB was able to decrease the hydrophobicity of a Teflon-like surface to the same extent as native RodA, while rRodA showed this ability to a lesser extent. Recombinant RodA and native...

Flowing Air-Water Cooled Slab Nd: Glass Laser

Science.gov (United States)

Lu, Baida; Cai, Bangwei; Liao, Y.; Xu, Shifa; Xin, Z.

1989-03-01

A zig-zag optical path slab geometry Nd: glass laser cooled through flowing air-water is developed by us. Theoretical studies on temperature distribution of slab and rod configurations in the unsteady state clarify the advantages of the slab geometry laser. The slab design and processing are also reported. In our experiments main laser output characteristics, e. g. laser efficiency, polarization, far-field divergence angle as well as resonator misalignment are investigated. The slab phosphate glass laser in combination with a crossed Porro-prism resonator demonstrates a good laser performance.

Selection of transverse modes in laser cavities containing waveguides and open parts

International Nuclear Information System (INIS)

Gurin, O V; Degtyarev, A V; Maslov, Vyacheslav A; Svich, V A; Tkachenko, V M; Topkov, A N

2001-01-01

The transverse modes of a submillimetre laser cavity that contains waveguides and open parts were studied theoretically and experimentally with the purpose of finding methods for mode selection. Two methods based on the filtering of the Fourier spectra of the waveguide modes and the use of their interference were substantiated numerically and realised in experiment. Special attention was paid to the mode selection in tunable lasers. Scaling laws allowing one to use the obtained results in a wide range of the cavity parameters and wavelengths are presented. (laser applications and other topics in quantum electronics)

Cone dystrophy with "supernormal" rod ERG: psychophysical testing shows comparable rod and cone temporal sensitivity losses with no gain in rod function.

Science.gov (United States)

Stockman, Andrew; Henning, G Bruce; Michaelides, Michel; Moore, Anthony T; Webster, Andrew R; Cammack, Jocelyn; Ripamonti, Caterina

2014-02-10

We report a psychophysical investigation of 5 observers with the retinal disorder "cone dystrophy with supernormal rod ERG," caused by mutations in the gene KCNV2 that encodes a voltage-gated potassium channel found in rod and cone photoreceptors. We compared losses for rod- and for cone-mediated vision to further investigate the disorder and to assess whether the supernormal ERG is associated with any visual benefit. L-cone, S-cone, and rod temporal acuity (critical flicker fusion frequency) were measured as a function of target irradiance; L-cone temporal contrast sensitivity was measured as a function of temporal frequency. Temporal acuity measures revealed that losses for vision mediated by rods, S-cones, and L-cones are roughly equivalent. Further, the gain in rod function implied by the supernormal ERG provides no apparent benefit to near-threshold rod-mediated visual performance. The L-cone temporal contrast sensitivity function in affected observers was similar in shape to the mean normal function but only after the mean function was compressed by halving the logarithmic sensitivities. The name of this disorder is potentially misleading because the comparable losses found across rod and cone vision suggest that the disorder is a generalized cone-rod dystrophy. Temporal acuity and temporal contrast sensitivity measures are broadly consistent with the defect in the voltage-gated potassium channel producing a nonlinear distortion of the photoreceptor response but after otherwise normal transduction processes.

Automation of the positioning of a laser anemometer flow rate measurement bench; Automatisation du positionnement d`un banc de mesure de vitesses d`ecoulements par anemometrie laser

Energy Technology Data Exchange (ETDEWEB)

Gobillot, G

1998-01-24

The laser anemometry technique is commonly used by the Core Hydraulics Laboratory of the CEA for the determination of the field of flow rates inside fuel rod bundles. The adjustment of measurement point coordinates represents an important part of the velocimetry campaign. In order to increase the number of measurements and the preciseness of the positioning operation, the automation of these preliminary tasks was decided. This work describes first the principle of Doppler laser velocimetry, the components of the measurement system and their functioning conditions. Then, the existing software for tuning and measurement is presented. A new software, called PAMELA, for the automatic positioning of the laser anemometer using a moving table with 5 degrees of freedom, has been developed and tested. This software, written with the LabView language, advises the operator, drives the bench and executes the tunings with a greater precision than manually. (J.S.) 16 refs.

Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

International Nuclear Information System (INIS)

Sepold, L.

1995-06-01

The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B 4 C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.) [de

Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

International Nuclear Information System (INIS)

Caraghiaur, Diana; Anglart, Henryk; Frid, Wiktor

2009-01-01

This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.

Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

Energy Technology Data Exchange (ETDEWEB)

Caraghiaur, Diana [Royal Institute of Technology, Division of Nuclear Reactor Technology, Department of Physics, School of Engineering Sciences, AlbaNova University Center, SE-106 91 Stockholm (Sweden)], E-mail: dianac@kth.se; Anglart, Henryk [Royal Institute of Technology, Division of Nuclear Reactor Technology, Department of Physics, School of Engineering Sciences, AlbaNova University Center, SE-106 91 Stockholm (Sweden); Frid, Wiktor [Swedish Radiation Safety Authority, Reactor Technology and Structural Integrity, SE-171 16 Stockholm (Sweden)

2009-10-15

This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.

Development of a Fiber Laser Welding Equipment for the LVDT Manufacturing

International Nuclear Information System (INIS)

Kim, S. S.; Lee, C. Y.; Park, S. J.; Shin, Y. T.; Cho, M. S.; Choo, K. N.; Kim, B. G.; Kang, Y. H.

2010-12-01

The instrumented fuel irradiation test at a research reactor is needed to evaluate the performance of the developed nuclear fuel. A LVDT(Linear Variable Differential Transformer) can be designed to measure the pressure level and elongation during the irradiation test by using various metals and MI cables. LVDT's parts were composed of MI cables and Inconel 600 materials. Laser welding technology is widely used to fabricate some products of nuclear fuel in the nuclear industry. Especially, micro-welding using Fiber laser is one of the key technologies to be developed to fabricate precise products of fuel irradiation test. We also have to secure micro-welding technology to perform various instrumentations for fuel irradiation test. Micro-welding technology was adopted to seal between seal tubes and MI cables with thickness of 0.15 mm. The soundness of welding area has to be confirmed to prevent fission gas of the fuel from leaking out of the fuel rods during the fuel irradiation test. In this report, fundamental data for welding technology using Fiber laser was proposed to seal Inconel 600 sheaths of LVDT instrumented capsules for the irradiation test. Moreover, It is expected that the use of fuel irradiation tests will be revitalized by the self-development of LVDT's parts using the Fiber laser welding technology

Improving solar-pumped laser efficiency by a ring-array concentrator

Science.gov (United States)

Tibúrcio, Bruno D.; Liang, Dawei; Almeida, Joana; Matos, Rodrigo; Vistas, Cláudia R.

2018-01-01

We report here a compact pumping scheme for achieving large improvement in collection and conversion efficiency of a Nd:YAG solar-pumped laser by an innovative ring-array solar concentrator. An aspheric fused silica lens was used to further concentrate the solar radiation from the focal region of the 1.5-m-diameter ring-array concentrator to a 5.0-mm-diameter, 20-mm-length Nd:YAG single-crystal rod within a conical-shaped pump cavity, enabling multipass pumping to the laser rod. 67.3-W continuous-wave solar laser power was numerically calculated, corresponding to 38.2-W / m2 solar laser collection efficiency, being 1.22 and 1.27 times more than the state-of-the-art records by both heliostat-parabolic mirror and Fresnel lens solar laser systems, respectively. 4.0% conversion efficiency and 0.021-W brightness figure of merit were also numerically obtained, corresponding to 1.25 and 1.62 times enhancement over the previous records, respectively. The influence of tracking error on solar laser output power was also analyzed.

In-pile experiments on fuel rod behavior during a LOCA

International Nuclear Information System (INIS)

Karb, E.; Pruessmann, M.; Sepold, L.

1980-05-01

This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

Control rod drive for vertical movement

International Nuclear Information System (INIS)

Suskov, I.I.; Gorjunov, V.S.; Zajcev, B.I.; Derevjankin, N.E.; Petrov, V.A.; Istomin, S.D.; Kovalencik, D.I.; Archipov, E.A.; Serebrjakov, V.I.; Kacalin, V.S.

1982-01-01

The control of the rod repositioning gear unit and the control unit of the profile grab of the control rod drive for the alkali metal-cooled fast breeder reactor is achieved by an electromotor being arranged outside the hermetic drive casing. The guide tube is directly repositioned by the rod repositioning gear unit. Coupling control of the drive with the control rod is done in the lower operative position of the control rod and that because of the interaction of the tie rod arranged on the spring-mounted control rod with the induction transmitter for the lower position of the control rod. In the transfer position the rod is fixed within the guide tube. (orig.)

Cone rod dystrophies

Science.gov (United States)

Hamel, Christian P

2007-01-01

Cone rod dystrophies (CRDs) (prevalence 1/40,000) are inherited retinal dystrophies that belong to the group of pigmentary retinopathies. CRDs are characterized by retinal pigment deposits visible on fundus examination, predominantly localized to the macular region. In contrast to typical retinitis pigmentosa (RP), also called the rod cone dystrophies (RCDs) resulting from the primary loss in rod photoreceptors and later followed by the secondary loss in cone photoreceptors, CRDs reflect the opposite sequence of events. CRD is characterized by primary cone involvement, or, sometimes, by concomitant loss of both cones and rods that explains the predominant symptoms of CRDs: decreased visual acuity, color vision defects, photoaversion and decreased sensitivity in the central visual field, later followed by progressive loss in peripheral vision and night blindness. The clinical course of CRDs is generally more severe and rapid than that of RCDs, leading to earlier legal blindness and disability. At end stage, however, CRDs do not differ from RCDs. CRDs are most frequently non syndromic, but they may also be part of several syndromes, such as Bardet Biedl syndrome and Spinocerebellar Ataxia Type 7 (SCA7). Non syndromic CRDs are genetically heterogeneous (ten cloned genes and three loci have been identified so far). The four major causative genes involved in the pathogenesis of CRDs are ABCA4 (which causes Stargardt disease and also 30 to 60% of autosomal recessive CRDs), CRX and GUCY2D (which are responsible for many reported cases of autosomal dominant CRDs), and RPGR (which causes about 2/3 of X-linked RP and also an undetermined percentage of X-linked CRDs). It is likely that highly deleterious mutations in genes that otherwise cause RP or macular dystrophy may also lead to CRDs. The diagnosis of CRDs is based on clinical history, fundus examination and electroretinogram. Molecular diagnosis can be made for some genes, genetic counseling is always advised. Currently

Cone rod dystrophies

Directory of Open Access Journals (Sweden)

Hamel Christian P

2007-02-01

Full Text Available Abstract Cone rod dystrophies (CRDs (prevalence 1/40,000 are inherited retinal dystrophies that belong to the group of pigmentary retinopathies. CRDs are characterized by retinal pigment deposits visible on fundus examination, predominantly localized to the macular region. In contrast to typical retinitis pigmentosa (RP, also called the rod cone dystrophies (RCDs resulting from the primary loss in rod photoreceptors and later followed by the secondary loss in cone photoreceptors, CRDs reflect the opposite sequence of events. CRD is characterized by primary cone involvement, or, sometimes, by concomitant loss of both cones and rods that explains the predominant symptoms of CRDs: decreased visual acuity, color vision defects, photoaversion and decreased sensitivity in the central visual field, later followed by progressive loss in peripheral vision and night blindness. The clinical course of CRDs is generally more severe and rapid than that of RCDs, leading to earlier legal blindness and disability. At end stage, however, CRDs do not differ from RCDs. CRDs are most frequently non syndromic, but they may also be part of several syndromes, such as Bardet Biedl syndrome and Spinocerebellar Ataxia Type 7 (SCA7. Non syndromic CRDs are genetically heterogeneous (ten cloned genes and three loci have been identified so far. The four major causative genes involved in the pathogenesis of CRDs are ABCA4 (which causes Stargardt disease and also 30 to 60% of autosomal recessive CRDs, CRX and GUCY2D (which are responsible for many reported cases of autosomal dominant CRDs, and RPGR (which causes about 2/3 of X-linked RP and also an undetermined percentage of X-linked CRDs. It is likely that highly deleterious mutations in genes that otherwise cause RP or macular dystrophy may also lead to CRDs. The diagnosis of CRDs is based on clinical history, fundus examination and electroretinogram. Molecular diagnosis can be made for some genes, genetic counseling is

Safety rod driving device

International Nuclear Information System (INIS)

Murakami, Kiyonobu; Kurosaki, Akira.

1988-01-01

Purpose: To rapidly insert safety rods for a criticality experiment device into a reactor core container to stop the criticality reaction thereby prevent reactivity accidents. Constitution: A cylinder device having a safety rod as a cylinder rod attached with a piston at one end is constituted. The piston is elevated by pressurized air and attracted and fixed by an electromagnet which is a stationary device disposed at the upper portion of the cylinder. If the current supply to the electromagnet is disconnected, the safety rod constituting the cylinder rod is fallen together with the piston to the lower portion of the cylinder. Since the cylinder rod driving device has neither electrical motor nor driving screw as in the conventional device, necessary space can be reduced and the weight is decreased. In addition, since the inside of the nuclear reactor can easily be shielded completely from the external atmosphere, leakage of radioactive materials can be prevented. (Horiuchi, T.)

Control Rod Reactivity Measurements in the Aagesta Reactor with the Pulsed Neutron Method

Energy Technology Data Exchange (ETDEWEB)

Bjoereus, K

1969-07-01

An extensive series of control rod measurements was made in the Aagesta reactor during the low power experimental period following the first criticality. This report describes the part of these investigations made with the pulsed neutron method, comprising nearly 300 measurements. The main objective was the determination of control rod reactivity worths for different rods and groups of rods, but some supplementary measurements were also made, e.g. a determination of the prompt neutron decay constant for the delayed critical condition and four different cores. The cores consisted of 20, 32, 68, and 140 fuel elements respectively, and measurements were made at room temperature and with the moderator level close to critical for each core, and for the 140-element core also with full moderator height and at the temperatures 140 deg C and 215 deg C. Both fully and partly inserted control rod groups were investigated. The measurements at critical water level give directly the control rod reactivity worths, whereas those with full water height give the shut-down reactivity. A comparison was made between measured reactivity worths for a number of rod groups and those calculated with the HETERO code. The prompt neutron decay constant at delayed criticality {alpha}{sub 0}={beta}/l, for the full core at 215 deg C was found to be 9.60 {+-} 0.30/sec, corresponding to l = 0.76 {+-} 0.02 msec. The shut-down reactivity with 16 coarse control rods in pos. A-D 22, 40-04, 44, 26 is -5% at 25 deg C and -13% at 215 deg C. The relative error is usually around 8% in the reactivity worths, originating mainly from the higher harmonics content in the measured curves.

Numerical experiment designs. Study of vibratory behaviour of PWR'S control rod clusters

International Nuclear Information System (INIS)

Bosselut, D.; Soulier, B.; Regnier, G.

1997-01-01

The application of Experiment Design method to Finite Element Model (FEM) calculations is an original way of performing parametric studies. It has been used at EDF to simulate on a large parametric domain the vibrations of PWR's control rod cluster and to analyse the rod wear process. In the first part the FEM and the location of excitation sources are described. The calculated values are: rod displacement in the guiding cards, shock forces on the guiding cards and the wear power produced. In the second part, the computed Experiment Domain is described. This method approaches the response surface by a second degree polynomial. The retained model is composed for every parameters of all linear, quadratic and interaction terms (26 coefficients). In all, 34 polynomials have been built to approach the effective shock forces and the mean wear power at each of the 17 guiding points. In the third part the building of the computer Experiment Design is detailed: by Doehlert design adaptation to take into account a qualitative parameter, design optimization by adding four well chosen experiments and finally, design extension by passing from 4 to 6 parameters. In the last part, all the information deduced from application of this method are presented. The influence of parameters on calculated effective shock forces has been determined along rods and response surface have been easily approximated. The systematism and closeness of Experiment Design technique is underlined. Easy simulation of all the response domain by polynomial approach allows comparison with experimental results. (authors)

5.4W cladding-pumped Nd:YAG silica fiber laser

OpenAIRE

Yoo, S.; Webb, A.S.; Standish, R.J.; May-Smith, T.C.; Sahu, J.K.

2012-01-01

We report on the spectroscopy and laser characteristics of Nd-doped fiber, fabricated by rod-in-tube from Nd:YAG as a core material with silica cladding. A cladding-pumped CW laser operation at 1058nm with 52% slope-efficiency is demonstrated.

Testing device for control rod drives

International Nuclear Information System (INIS)

Hayakawa, Toshifumi.

1992-01-01

A testing device for control rod drives comprises a logic measuring means for measuring an output signal from a control rod drive logic generation circuit, a control means for judging the operation state of a control rod and a man machine interface means for outputting the result of the judgement. A driving instruction outputted from the control rod operation device is always monitored by the control means, and if the operation instruction is stopped, a testing signal is outputted to the control rod control device to simulate a control rod operation. In this case, the output signal of the control rod drive logic generation circuit is held in a control rod drive memory means and intaken into a logic analysis means for measurement and an abnormality is judged by the control means. The stopping of the control rod drive instruction is monitored and the operation abnormality of the control rod is judged, to mitigate the burden of an operator. Further, the operation of the control rod drive logic generation circuit can be confirmed even during a nuclear plant operation by holding the control rod drive instruction thereby enabling to improve maintenance efficiency. (N.H.)

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 9 discusses the following topics: Integrated System Normal Operations Test Results and Analysis Report; Integrated System Off-Normal Operations Test Results and Analysis Report; and Integrated System Maintenance Operations Test Results and Analysis Report

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 1 discusses the following topics: the background of the project; test program description; summary of tests and test results; problem evaluation; functional requirements confirmation; recommendations; and completed test documentation for tests performed in Phase 3

Prototypical Rod Construction Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 3 discusses the following topics: Downender Test Results and Analysis Report; NFBC Canister Upender Test Results and Analysis Report; Fuel Assembly Handling Fixture Test Results and Analysis Report; and Fuel Canister Upender Test Results and Analysis Report

Automation of the positioning of a laser anemometer flow rate measurement bench

International Nuclear Information System (INIS)

Gobillot, G.

1998-01-01

The laser anemometry technique is commonly used by the Core Hydraulics Laboratory of the CEA for the determination of the field of flow rates inside fuel rod bundles. The adjustment of measurement point coordinates represents an important part of the velocimetry campaign. In order to increase the number of measurements and the preciseness of the positioning operation, the automation of these preliminary tasks was decided. This work describes first the principle of Doppler laser velocimetry, the components of the measurement system and their functioning conditions. Then, the existing software for tuning and measurement is presented. A new software, called PAMELA, for the automatic positioning of the laser anemometer using a moving table with 5 degrees of freedom, has been developed and tested. This software, written with the LabView language, advises the operator, drives the bench and executes the tunings with a greater precision than manually. (J.S.)

Failed fuel rod detector

Energy Technology Data Exchange (ETDEWEB)

Uchida, Katsuya; Matsuda, Yasuhiko

1984-05-02

The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

RodPilotR - The Innovative and Cost-Effective Digital Control Rod Drive Control System for PWRs

International Nuclear Information System (INIS)

Baron, Clemens

2008-01-01

With RodPilot, AREVA NP offers an innovative and cost-effective system for controlling control rods in Pressurized Water Reactors. RodPilot controls the three operating coils of the control rod drive mechanism (lift, moveable gripper and stationary gripper coil). The rods are inserted into or withdrawn from the core as required by the Reactor Control System. The system combines modern components, state-of-the-art logic and a proven electronic control rod drive control principle to provide enhanced reliability and lower maintenance costs. (author)

Integrated control rod monitoring device

International Nuclear Information System (INIS)

Saito, Katsuhiro

1997-01-01

The present invention provides a device in which an entire control rod driving time measuring device and a control rod position support device in a reactor building and a central control chamber are integrated systematically to save hardwares such as a signal input/output device and signal cables between boards. Namely, (1) functions of the entire control rod driving time measuring device for monitoring control rods which control the reactor power and a control rod position indication device are integrated into one identical system. Then, the entire devices can be made compact by the integration of the functions. (2) The functions of the entire control rod driving time measuring device and the control rod position indication device are integrated in a central operation board and a board in the site. Then, the place for the installation of them can be used in common in any of the cases. (3) The functions of the entire control rod driving time measuring device and the control rod position indication device are integrated to one identical system to save hardware to be used. Then, signal input/output devices and drift branching panel boards in the site and the central operation board can be saved, and cables for connecting both of the boards is no more necessary. (I.S.)

Rod drive and latching mechanism

International Nuclear Information System (INIS)

Veronesi, L.; Sherwood, D.G.

1982-01-01

Hydraulic drive and latching mechanisms for driving reactivity control mechanisms in nuclear reactors are described. Preferably, the pressurized reactor coolant is utilized to raise the drive rod into contact with and to pivot the latching mechanism so as to allow the drive rod to pass the latching mechanism. The pressure in the housing may then be equalized which allows the drive rod to move downwardly into contact with the latching mechanism but to hold the shaft in a raised position with respect to the reactor core. Once again, the reactor coolant pressure may be utilized to raise the drive rod and thus pivot the latching mechanism so that the drive rod passes above the latching mechanism. Again, the mechanism pressure can be equalized which allows the drive rod to fall and pass by the latching mechanism so that the drive rod approaches the reactor core. (author)

Reactivity shimming in FFTF with safety rods

International Nuclear Information System (INIS)

Bennett, C.L.

1986-01-01

The Fast Flux Test Facility (FFTF) reactor demonstrated operation at a 65% annual capacity factor in 1984 and a 70% annual capacity factor in 1985. This was achieved in part by extending the cycle lengths beyond the originally designed 100-day cycles. The advent of extended cycle lengths at the FFTF reactor led to the development of a contingency plan to shim some of the installed excess reactivity at the beginning of the cycle with the primary control rod bank (safety rods) in order to maintain the minimum required shutdown margin in the secondary shutdown system. This paper describes both the implementation of this plan in terms of the licensing aspects and the actual use of primary shim during cycles 5 and 6 operation at FFTF

Control rods

International Nuclear Information System (INIS)

Koga, Isao; Masuoka, Ryuzo.

1979-01-01

Purpose: To prevent fuel element failures during power conditioning by removing liquid absorbents in poison tubes of control rods in a fast power up step and extracting control rods to slightly increase power in a medium power up step. Constitution: A plurality of poison tubes are disposed in a coaxial or plate-like arrangement and divided into a region capable of compensating the reactivity from the initial state at low temperature to 40% power operation and a region capable of compensating the reactivity in the power up above 40% power operation. Soluble poisons are used as absorbers in the poison tubes corresponding to above 40% power operation and they are adapted to be removed independently from the driving of control rods. The poison tubes filled with the soluble absorbers are responsible for the changes in the reactivity from the initial state at low temperature to the medium power region and the reactivity control is conducted by the elimination of liquid absorbers from the poison tubes. In the succeeding slight power up region above the medium power, power up is proceeding by extracting the control rods having remaining poison tubes filled with solid or liquid absorbers. (Seki, T.)

Nondestructive examination of irradiated fuel rods by pulsed eddy current techniques

International Nuclear Information System (INIS)

Francis, W.C.; Quapp, W.J.; Martin, M.R.; Gibson, G.W.

1976-02-01

A number of fuel rods and unfueled zircaloy cladding tubes which had been irradiated in the Saxton reactor have undergone extensive nondestructive and corroborative destructive examinations by Aerojet Nuclear Company as part of the Water Reactor Safety Research Program, Irradiation Effects Test Series. This report discusses the pulsed eddy current (PEC) nondestructive examinations on the fuel rods and tubing and the metallography results on two fuel rods and one irradiated zircaloy tube. The PEC equipment, designed jointly by Argonne National Laboratory and Aerojet, performed very satisfactorily the functions of diameter, profile, and wall thickness measurements and OD and ID surface defect detection. The destructive examination provided reasonably good confirmation of ''defects'' detected in the nondestructive examination

Assessment of US NRC fuel rod behavior codes to extended burnup

International Nuclear Information System (INIS)

Laats, E.T.; Croucher, D.W.; Haggag, F.M.

1982-01-01

The purpose of this paper is to report the status of assessing the capabilities of the NRC fuel rod performance codes for calculating extended burnup rod behavior. As part of this effort, a large spectrum of fuel rod behavior phenomena was examined, and the phenomena deemed as being influential during extended burnup operation were identified. Then, the experiment data base addressing these identified phenomena was examined for availability and completeness at extended burnups. Calculational capabilities of the NRC's steady state FRAPCON-2 and transient FRAP-T6 fuel rod behavior codes were examined for each of the identified phenomenon. Parameters calculated by the codes were compared with the available data base, and judgments were made regarding model performance. Overall, the FRAPCON-2 code was found to be moderately well assessed to extended burnups, but the FRAP-T6 code cannot be adequately assessed until more transient high burnup data are available

Drying of encapsulated parts (nuclear fuel rods) in applying vacuum, by introducing dehydratings, vacuum, and filling with an inert gas

International Nuclear Information System (INIS)

Johnson, C.R.

1976-01-01

This invention concerns a decontamination technique, in particular a process and equipment for extracting the water contained in fuel rods and other similar components of a nuclear reactor. The extraction of the contaminants contained in the fuel rods is carried out by a standard method by drilling a small hole in the surface of the cladding and applying a vacuum to bleed the rod of its impurities (moisture and gas). The invention consists for example in applying a vacuum at the hole drilled in the cladding to extract the contaminants and introducing spirit into the rod through the same orifice. The spirit absorbs the remaining liquid and other impurities. The spirit charged with the impurities is then pumped out by the same aperture by means of a regulated atmosphere inside a closed receptacle. This receptacle is then filled with an inert gas cooled to ambient temperature. The rods are then pressurised and the small orifice is sealed [fr

1.8kW laser diode pumped YAG laser; Shutsuryoku 1.8kW no handotai laser reiki YAG laser

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-03-01

Toshiba Corporation, as a participant in Ministry of International Trade and Industry`s `photon measurement and processing technology project` since August, 1997, is engaged in the development of an energy-efficient LD (laser diode) pumped semiconductor YAG (yttrium-aluminum-garnet) laser device to be used for welding and cutting. It is a 5-year project and the goal is a mean output of 10kW and efficiency of 20%. In this article, a simulation program is developed which carries out calculation about element technology items such as the tracking of the beam from the pumping LD and the excitation distribution, temperature distribution, thermal stress distribution, etc., in the YAG rod. An oscillator is constructed, based on the results of the simulation, and it exhibits a world-high class continuous laser performance of a 1.8kW output and 13% efficiency. The record of 13% efficiency is five times higher than that achieved by the conventional lamp-driven YAG laser device. (translated by NEDO)

The structure of single-phase turbulent flows through closely spaced rod arrays

International Nuclear Information System (INIS)

Hooper, J.D.; Rehme, K.

1983-02-01

The axial and azimuthal turbulence intensity in the rod gap region has been shown, for developed single-phase turbulent flow through parallel rod arrays, to strongly increase with decreasing rod spacing. Two array geometries are reported, one constructed from a rectangular cross-section duct containing four rods and spaced at five p/d or w/d ratios. The second test section, constructed from six rods set in a regular square-pitch array, represented the interior flow region of a large array. The mean axial velocity, wall shear stress variation and axial pressure distribution were measured, together with hot-wire anemometer measurements of the Reynolds stresses. No significant non-zero secondary flow components were detected, using techniques capable of resolving secondary flow velocities to 1% of the local axial velocity. For the lowest p/d ratio of 1.036, cross-correlation measurements showed the presence of an energetic periodic azimuthal turbulent velocity component, correlated over a significant part of the flow area. The negligible contribution of secondary flows to the axial momentum balance, and the large azimuthal turbulent velocity component in the rod gap area, suggest a different mechanism than Reynolds stress gradient driven secondary flows for the turbulent transport process in the rod gap. (orig.) [de

Fuel rod technology

International Nuclear Information System (INIS)

Bezold, H.; Romeiser, H.J.

1979-07-01

By extensive mechanization and automation of the fuel rod production, also at increasing production numbers, an efficient production shall be secured, simultaneously corresponding to the high quality standard of the fuel rods. The works done up to now concentrated on the lay out of a rough concept for a mechanized production course. Detail-studies were made for the problems of fuel rod humidity, filling and resistance welding. Further promotion of this project and thus further report will be stopped, since the main point of these works is the production technique. (orig.) [de

Temperature studies of the TileCal ROD G-Links for the validation of the air-cooling system

CERN Document Server

Valero, A; Abdallah, J; Castillo, V; Cuenca, C; Ferrer, A; Fullana, E; González, V; Higón, E; Munar, A; Poveda, J; Salvachúa, B; Sanchis, E; Solans, C; Torres, J; Valls, J A

2007-01-01

In this paper we show the results of the temperature studies performed on the TileCal ROD G-Links in order to validate the air-cooling system. In the first part of the note we present results on the characterization tests of the temperature monitor system for the G-Link chips of the TileCal ROD motherboard, performed at IFIC-Valencia. We report on the performance of the temperature behavior system and some cooling studies of a single ROD motherboard. We conclude that the present system can be successfully used to online monitor the temperature of the ROD G-Links. In the second part we show the results of the studies performed with multiple RODs in a standard 9U VME crate in the laboratory at IFIC, and in their final location in the ATLAS cavern. We conclude that the air-cooling provided by the standard VME crate fans is enough to keep the temperature of the G-Links well within specifications.

Measurements of two-phase flow patterns in a 4 x 4 rod bundle

International Nuclear Information System (INIS)

Akio tomiyama; Akira Sou; Shigeo Hosokawa; Masato Mitsuhashi; Kohei Noda; Yasushi Tsubo; Kaichiro Mishima; Yoshiro Kudo

2005-01-01

Air-water two-phase flow patterns in a 4 x 4 square lattice rod bundle consisting of an acrylic channel box of 68 mm in width and transparent rods of 12 mm in diameter were measured by utilizing FEP (fluorinated ethylene propylene) tubes for the rods. The FEP possesses the same refractive index with water, and therefore, whole flow patterns in the bundle and local flow patterns in subchannels were visualized with little optical distortion. In addition to the visualization, transmission rates of laser beam from one rod to its opponent rod and two-point correlation coefficients of phase indicator functions were measured to examine the feasibility of objective identification of flow patterns in subchannels. The ranges of liquid and gas volume fluxes, JL and JG, were 0.1 < JL < 2.0 m/s and 0.04 < JG < 8.85 m/s, respectively. As a result, the following conclusions were obtained: (1) slug flow pattern does not appear in the rod bundle and bubbly flow would directly transit to churn flow, (2) the measured boundary between bubbly and churn flows is close to the boundary between bubbly and slug flows given by Mishima and Ishii's flow pattern transition model, (3) critical void fraction causing bubbly to churn flow transition depends on a subchannel, i.e., about 0.3 for inner subchannels, about 0.2 for side subchannels and about 0.1 for corner subchannels, and (4) the two-point correlation coefficient of phase indicator functions for two inner subchannels shows a steep increase at the bubbly to churn flow transition, which, in turn, means that the two-point correlation is an appropriate indicator for detecting this transition. (authors)

Digital control rod blocking monitor

International Nuclear Information System (INIS)

Funayama, Yoshio.

1996-01-01

The present invention system is used for monitoring of a power region of a reactor, and used for monitoring of simultaneous withdrawal of a plurality of control rods without increasing the size or complicating the system. Namely, the system processes signals from a neutron flux detectors at the periphery of control rods controlled for withdrawal. As a result of the processing, the digital monitoring system generates an alarm when the reactor power at the periphery of the control rods fluctuates exceeding an allowable range. In the system, a control rod information forming means prepares frame data comprising front data, positions of the control rods to be withdrawn, frame numbers and completion data. A serial data transmitting means transmits the frame data successively as repeating frame data rows. A control rod information receiving means takes up the frame data of each of control rods to be withdrawn from the transmitted frame data rows. Since the system of the present invention can monitor the withdrawal of a plurality of control rods simultaneously without increasing the size or complicating the system, cost can be saved and the maintenance can be improved. (I.S.)

Liquid momentum removal using rod arrays applied to the HYLIFE ICF reactor

International Nuclear Information System (INIS)

Raffray, A.R.; Hoffman, M.A.

1986-01-01

This research relates to the multiple liquid-lithium-jet blanket concept for the HYLIFE inertial-confinement fusion (ICF) reactor. The fusion micro-explosion would result in part of the liquid lithium being propelled towards the vacuum chamber wall where the resulting impact would cause high peak stresses. In an attempt to reduce these peak stresses, it was proposed to set up an array of bars between the vacuum vessel first wall and the liquid jets so that part of the liquid momentum would be removed as the liquid passed through the bars. A series of small-scale scoping experiments were run to obtain a preliminary evaluation of the effectiveness of such rod arrays in removing momentum from impinging liquid slugs. The impact force of an unconfined cylindrical water jet on in-line and staggered rod arrays was measured. The results indicate that the fraction of momentum removed from liquid slugs could probably exceed 18% for a staggered rod arrangement in the HYLIFE reactor

A NEW ELECTRONIC BOARD TO DRIVE THE LASER CALIBRATION SYSTEM OF THE ATLAS HADRON CALORIMETER

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00086824; The ATLAS collaboration

2016-01-01

The LASER calibration system of the ATLAS hadron calorimeter aims at monitoring the ~10000 PMTs of the TileCal. The LASER light injected in the PMTs is measured by sets of photodiodes at several stages of the optical path. The monitoring of the photodiodes is performed by a redundant internal calibration system using an LED, a radioactive source, and a charge injection system. The LASer Calibration Rod (LASCAR) electronics card is a major component of the LASER calibration scheme. Housed in a VME crate, its main components include a charge ADC, a TTCRx, a HOLA part, an interface to control the LASER, and a charge injection system. The 13 bits ADC is a 2000pc full-scale converter that processes up to 16 signals stemming from 11 photodiodes, 2 PMTs, and 3 charge injection channels. Two gains are used (x1 and x4) to increase the dynamic range and avoid a saturation of the LASER signal for high intensities. The TTCRx chip (designed by CERN) retrieves LHC signals to synchronize the LASCAR card with the collider. T...

Control rods

International Nuclear Information System (INIS)

Hirukawa, Koji.

1979-01-01

Purpose: To ensure the fuel safety by constituting a control rod with a plurality of poison bodies suspended in a cross-like section and shorter length poison bodies made movable and engageable in the gap between each of the above poison bodies thereby maintaining the function of the shorter length poison constant. Constitution: Cross-like supports are secured to the upper and lower parts of a driving shaft journaled in a sheath and poison bodies composed of neutron absorber poisons of a large thermal neutron absorption cross section and neutron absorber poison tubes for containing them are suspended from the supports. A movable cross-like support is mounted slidably at its base to the lower part of the driving shaft and poison bodies shorter than the above poison bodies and composed of neutron absorber poisons having a greater absorption cross section at the neutron energy region higher than thermal neutron region and neutron poison tubes for containing them are suspended to the movable support at the position capable of engaging in the gap between each of the poison bodies. (Kawakami, Y.)

The BWR Hybrid 4 control rod

International Nuclear Information System (INIS)

Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

1988-01-01

The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

Burnable poison rod

International Nuclear Information System (INIS)

Natsume, Tomohiro.

1988-01-01

Purpose: To decrease the effect of water elimination and the effect of burn-up residue boron, thereby reduce the effect of burnable poison rods as the neutron poisons at the final stage of reactor core lifetime. Constitution: In a burnable poison rod according to the present invention, a hollow burnable poison material is filled in an external fuel can, an inner fuel can mounted with a carbon rod is inserted to the hollow portion of the burnable poison material and helium gases are charged in the outer fuel can. In such a burnable poison rod, the reactivity worths after the burning are reduced to one-half as compared with the conventional case. Accordingly, since the effect of the burnable poison as the neutron poisons is reduced at the final stage of the reactor core of lifetime, the excess reactivity of the reactor core is increased. (Horiuchi, T.)

Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

International Nuclear Information System (INIS)

Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

1978-09-01

In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

Components inspection of Monju, a sodium bonded type control rod

International Nuclear Information System (INIS)

Harada, Kiyoshi; Matsushita, Yuichi; Lee, Chunchan; Abe, Hideaki; Watahiki, Naohisa

2002-03-01

This Report addresses a result of a sodium test conducted on components of a Double Poral Filter Sodium Bonded Type Control Rod that is expected to be a next generation, long life Control Rod. Upper and lower Poral Filter Sodium Bonded Type Control Rod components were mocked up to conduct a sodium test. During the test, sodium chargeability, formation of Gas Plenum at the upper part of the components, sodium drain-ability and NaOH clean-ability were recognized under actual plant condition. The following are results obtained: (1) Sodium Chargeability at Control Rod Insertion to EVST. Sodium was charged into the components when the mocked-up was inserted in sodium of 190degC, with insertion speed of 6 m/min which is an actual insertion speed to EVST. (2) Formation of Upper Gas Plenum by Helium Gas generated in Control Rod Components Gas Plenum formation within deviation of 9% was confirmed by releasing helium gas into the mocked-up which is immersed in sodium of 620degC and 190degC. Length of Gas Plenum is confirmed to be retained in certain length even if helium gas is further released into formed Gas Plenum. (3) Sodium Drain-ability of Control Rod Components when Drawing from EVST. Drain-ability was confirmed to be sufficient and no sodium residue was found in the mocked-up when the mocked-up was drawn out from sodium of 190degC, with drawing speed of 6 m/min which is an actual drawing speed from EVST. (4) Clean-ability of NaOH Solution against Sodium Residue in Control Rod Components. Sodium and NaOH solution reacted calmly, however, clean-ability was not sufficient. When Sodium fully remained in Control Rod Components, it made circulation of NaOH solution not enough. (author)

Biomechanics of lumbar cortical screw-rod fixation versus pedicle screw-rod fixation with and without interbody support.

Science.gov (United States)

Perez-Orribo, Luis; Kalb, Samuel; Reyes, Phillip M; Chang, Steve W; Crawford, Neil R

2013-04-15

Seven different combinations of posterior screw fixation, with or without interbody support, were compared in vitro using nondestructive flexibility tests. To study the biomechanical behavior of a new cortical screw (CS) fixation construct relative to the traditional pedicle screw (PS) construct. The CS is an alternative to the PS for posterior fixation of the lumbar spine. The CS trajectory is more sagittally and cranially oriented than the PS, being anchored in the pars interarticularis. Like PS fixation, CS fixation uses interconnecting rods fastened with top-locking connectors. Stability after bilateral CS fixation was compared with stability after bilateral PS fixation in the setting of intact disc and with direct lateral interbody fixation (DLIF) or transforaminal lateral interbody fixation (TLIF) support. Standard nondestructive flexibility tests were performed in cadaveric lumbar specimens, allowing non-paired comparisons of specific conditions from 28 specimens (4 groups of 7) within a larger experiment of multiple hardware configurations. Condition tested and group from which results originated were as follows: (1) intact (all groups); (2) with L3-L4 bilateral PS-rods (group 1); (3) with bilateral CS-rods (group 2); (4) with DLIF (group 3); (5) with DLIF + CS-rods (group 4); (6) with DLIF + PS-rods (group 3); (7) with TLIF + CS-rods (group 2), and (8) with TLIF + PS-rods (group 2). To assess spinal stability, the mean range of motion, lax zone, and stiff zone at L3-L4 were compared during flexion-extension, lateral bending, and axial rotation. With intact disc, stability was equivalent after PS-rod and CS-rod fixation, except that PS-rod fixation was stiffer during axial rotation. With DLIF support, there was no significant difference in stability between PS-rod and CS-rod fixation. With TLIF support, PS-rod fixation was stiffer than CS-rod fixation during lateral bending. Bilateral CS-rod fixation provided about the same stability in cadaveric specimens

Freely suspended rod fall dampener, especially for control rod of liquid-cooled nuclear reactor

International Nuclear Information System (INIS)

Becvar, J.; Saroch, V.

1977-01-01

A shock absorber is described whose advantage is that the space required for the movement of the shock absorber in the operating travel of the system suspension rod-control rod bundle may be reduced. The design allows the automatic disconnection of the system and the removal of the suspension rod from the reactor without dismantling. The braking force reaction is transmitted to the structure above the core. The system fall energy is absorbed on the side of the suspension rod which has a bigger mass. (J.B.)

RodPilot{sup R} - The Innovative and Cost-Effective Digital Control Rod Drive Control System for PWRs

Energy Technology Data Exchange (ETDEWEB)

Baron, Clemens [AREVA NP GmbH, NLEE-G, Postfach 1199, 91001 Erlangen (Germany)

2008-07-01

With RodPilot, AREVA NP offers an innovative and cost-effective system for controlling control rods in Pressurized Water Reactors. RodPilot controls the three operating coils of the control rod drive mechanism (lift, moveable gripper and stationary gripper coil). The rods are inserted into or withdrawn from the core as required by the Reactor Control System. The system combines modern components, state-of-the-art logic and a proven electronic control rod drive control principle to provide enhanced reliability and lower maintenance costs. (author)

Tensile Properties Characterization of AlSi10Mg Parts Produced by Direct Metal Laser Sintering via Nested Effects Modeling.

Science.gov (United States)

Palumbo, Biagio; Del Re, Francesco; Martorelli, Massimo; Lanzotti, Antonio; Corrado, Pasquale

2017-02-08

A statistical approach for the characterization of Additive Manufacturing (AM) processes is presented in this paper. Design of Experiments (DOE) and ANalysis of VAriance (ANOVA), both based on Nested Effects Modeling (NEM) technique, are adopted to assess the effect of different laser exposure strategies on physical and mechanical properties of AlSi10Mg parts produced by Direct Metal Laser Sintering (DMLS). Due to the wide industrial interest in AM technologies in many different fields, it is extremely important to ensure high parts performances and productivity. For this aim, the present paper focuses on the evaluation of tensile properties of specimens built with different laser exposure strategies. Two optimal laser parameters settings, in terms of both process quality (part performances) and productivity (part build rate), are identified.

Control-rod driving mechanism

International Nuclear Information System (INIS)

Jodoi, Takashi.

1976-01-01

Purpose: To prevent falling of control rods due to malfunction. Constitution: The device of the present invention has a scram function in particular, and uses principally a fluid pressure as a scram accelerating means. The control rod is held by upper and lower holding devices, which are connected by a connecting mechanism. This connecting mechanism is designed to be detachable only at the lower limit of driving stroke of the control rod so that there occurs no erroneous scram resulting from careless disconnection of the connecting mechanism. Further, scramming operation due to own weight of the scram operating portion such as control rod driving shaft may be effected to increase freedom. (Kamimura, M.)

Time dependent temperature distribution in pulsed Ti:sapphire lasers

Science.gov (United States)

Buoncristiani, A. Martin; Byvik, Charles E.; Farrukh, Usamah O.

1988-01-01

An expression is derived for the time dependent temperature distribution in a finite solid state laser rod for an end-pumped beam of arbitrary shape. The specific case of end pumping by circular (constant) or Gaussian beam is described. The temperature profile for a single pump pulse and for repetitive pulse operation is discussed. The particular case of the temperature distribution in a pulsed titanium:sapphire rod is considered.

Recent Advances in the Study of Bipolar/Rod-Shaped Microglia and their Roles in Neurodegeneration

Directory of Open Access Journals (Sweden)

Ngan Pan Bennett Au

2017-05-01

Full Text Available Microglia are the resident immune cells of the central nervous system (CNS and they contribute to primary inflammatory responses following CNS injuries. The morphology of microglia is closely associated with their functional activities. Most previous research efforts have attempted to delineate the role of ramified and amoeboid microglia in the pathogenesis of neurodegenerative diseases. In addition to ramified and amoeboid microglia, bipolar/rod-shaped microglia were first described by Franz Nissl in 1899 and their presence in the brain was closely associated with the pathology of infectious diseases and sleeping disorders. However, studies relating to bipolar/rod-shaped microglia are very limited, largely due to the lack of appropriate in vitro and in vivo experimental models. Recent studies have reported the formation of bipolar/rod-shaped microglia trains in in vivo models of CNS injury, including diffuse brain injury, focal transient ischemia, optic nerve transection and laser-induced ocular hypertension (OHT. These bipolar/rod-shaped microglia formed end-to-end alignments in close proximity to the adjacent injured axons, but they showed no interactions with blood vessels or other types of glial cell. Recent studies have also reported on a highly reproducible in vitro culture model system to enrich bipolar/rod-shaped microglia that acts as a powerful tool with which to characterize this form of microglia. The molecular aspects of bipolar/rod-shaped microglia are of great interest in the field of CNS repair. This review article focuses on studies relating to the morphology and transformation of microglia into the bipolar/rod-shaped form, along with the differential gene expression and spatial distribution of bipolar/rod-shaped microglia in normal and pathological CNSs. The spatial arrangement of bipolar/rod-shaped microglia is crucial in the reorganization and remodeling of neuronal and synaptic circuitry following CNS injuries. Finally, we

REACTOR CONTROL ROD OPERATING SYSTEM

Science.gov (United States)

Miller, G.

1961-12-12

A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

Design of high-brightness TEM00-mode solar-pumped laser for renewable material processing

Science.gov (United States)

Liang, D.; Almeida, J.

2014-08-01

The conversion of sunlight into laser light by direct solar pumping is of ever-increasing importance because broadband, temporally constant, sunlight is converted into laser light, which can be a source of narrowband, collimated, rapidly pulsed, radiation with the possibility of obtaining extremely high brightness and intensity. Nonlinear processes, such as harmonic generation, might be used to obtain broad wavelength coverage, including the ultraviolet wavelengths, where the solar flux is very weak. The direct excitation of large lasers by sunlight offers the prospect of a drastic reduction in the cost of coherent optical radiation for high average power materials processing. This renewable laser has a large potential for many applications such as high-temperature materials processing, renewable magnesium-hydrogen energy cycle and so on. We propose here a scalable TEM00 mode solar laser pumping scheme, which is composed of four firststage 1.13 m diameter Fresnel lenses with its respective folding mirrors mounted on a two-axis automatic solar tracker. Concentrated solar power at the four focal spots of these Fresnel lenses are focused individually along a common 3.5 mm diameter, 70 mm length Nd:YAG rod via four pairs of second-stage fused-silica spherical lenses and third-stage 2D-CPCs (Compound Parabolic Concentrator), sitting just above the laser rod which is also double-pass pumped by four V-shaped pumping cavities. Distilled water cools both the rod and the concentrators. 15.4 W TEM00 solar laser power is numerically calculated, corresponding to 6.7 times enhancement in laser beam brightness.

Making effective use of rod pumping systems in coalbed methane applications

Energy Technology Data Exchange (ETDEWEB)

Crivello, A. [eProduction Solutions Inc., Kingwood, TX (United States)

2003-07-01

The advantages of optimizing coalbed methane (CBM) operations are increased production, reduced expenses, improved efficiency, and better inventory. The author discussed the CBM production cycle and the possible artificial lift options, including electric submersible pump (ESP), plunger lift, primary coolant pump (PCP), and reciprocating rod lift. The presentation focused on the rod lift, as it represents a low to moderate capital expenditure, has good system efficiency, an excellent fluid volume range, an excellent salvage value, excellent familiarity with equipment, and has readily available parts and service. The major disadvantage of the rod lift is that the fixed operating range does not adapt to changing reservoir characteristics. A comparison between the rod pump controller and the variable speed drive was presented. The well can be operated at or near the pumped off condition with variable speed drives with rod pumping intelligence. The author provided a closer examination of the variable frequency drive and the vector flux drive. The presentation also included a discussion of prime movers, drive and inclinometer, gearbox loading, rod load limiter, and dynamometer cards. Three case studies were presented: CSW1, CSW2, and CSW3. It was concluded that wells must be kept pumping, and that a Flux Vector Drive should be used along with an NEMA B motor and properly sized pumping unit and pump. tabs., figs.

Control rod velocity limiter

International Nuclear Information System (INIS)

Cearley, J.E.; Carruth, J.C.; Dixon, R.C.; Spencer, S.S.; Zuloaga, J.A. Jr.

1986-01-01

This patent describes a velocity control arrangement for a reciprocable, vertically oriented control rod for use in a nuclear reactor in a fluid medium, the control rod including a drive hub secured to and extending from one end therefrom. The control device comprises: a toroidally shaped control member spaced from and coaxially positioned around the hub and secured thereto by a plurality of spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the toroidal member spaced therefrom in coaxial position. The side of the control member toward the control rod has a smooth generally conical surface. The side of the control member away from the control rod is formed with a concave surface constituting a single annular groove. The device also comprises inner and outer annular vanes radially spaced from one another and spaced from the side of the control member away from the control rod and positioned coaxially around and spaced from the hub and secured thereto by spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the vanes. The vanes are angled toward the control member, the outer edge of the inner vane being closer to the control member and the inner edge of the outer vane being closer to the control member. When the control rod moves in the fluid in the direction toward the drive hub the vanes direct a flow of fluid turbulence which provides greater resistance to movement of the control rod in the direction toward the drive hub than in the other direction

183-W, M2 = 2.4 Yb:YAG Q -switched laser

International Nuclear Information System (INIS)

Honea, E.C.; Beach, R.J.; Mitchell, S.C.; Avizonis, P.V.

1999-01-01

We have fabricated a diode-array end-pumped Yb:YAG rod laser with output powers greater than 200thinspthinspW cw and 195thinspthinspW Q -switched at 5thinspthinspkHz. At an output power of 183thinspthinspW and a repetition rate of 5thinspthinspkHz, the beam quality was measured to be M 2 =2.4 . The laser design incorporates a hollow lens duct to concentrate the diode pump light for delivery to the end of the laser rod while maintaining access to the laser beam. This configuration provides increased flexibility for the resonator design and permits the use of birefringence compensation in the cavity to yield polarized output with increased efficiency. Using the recently described birefringence compensation method of Clarkson et al.thinspthinsp[in Conference on Lasers and Electro-Optics (Optical Society of America, Washington, D.C., 1998), paper CTuI3], we obtained 112thinspthinspW of cw power with a polarized beam of M 2 =3.2 . copyright 1999 Optical Society of America

Control rod position control device

International Nuclear Information System (INIS)

Ubukata, Shinji.

1997-01-01

The present invention provides a control rod position control device which stores data such as of position signals and driving control rod instruction before and after occurrence of abnormality in control for the control rod position for controlling reactor power and utilized the data effectively for investigating the cause of abnormality. Namely, a plurality of individual control devices have an operation mismatching detection circuit for outputting signals when difference is caused between a driving instruction given to the control rod position control device and the control rod driving means and signals from a detection means for detecting an actual moving amount. A general control device collectively controls the individual control devices. In addition, there is also disposed a position storing circuit for storing position signals at least before and after the occurrence of the control rod operation mismatching. With such procedures, the cause of the abnormality can be determined based on the position signals before and after the occurrence of control rod mismatching operation stored in the position storing circuit. Accordingly, the abnormality cause can be determined to conduct restoration in an early stage. (I.S.)

Tensile Properties Characterization of AlSi10Mg Parts Produced by Direct Metal Laser Sintering via Nested Effects Modeling

Directory of Open Access Journals (Sweden)

Biagio Palumbo

2017-02-01

Full Text Available A statistical approach for the characterization of Additive Manufacturing (AM processes is presented in this paper. Design of Experiments (DOE and ANalysis of VAriance (ANOVA, both based on Nested Effects Modeling (NEM technique, are adopted to assess the effect of different laser exposure strategies on physical and mechanical properties of AlSi10Mg parts produced by Direct Metal Laser Sintering (DMLS. Due to the wide industrial interest in AM technologies in many different fields, it is extremely important to ensure high parts performances and productivity. For this aim, the present paper focuses on the evaluation of tensile properties of specimens built with different laser exposure strategies. Two optimal laser parameters settings, in terms of both process quality (part performances and productivity (part build rate, are identified.

Numerical simulation of thermal loading produced by shaped high power laser onto engine parts

International Nuclear Information System (INIS)

Song Hongwei; Li Shaoxia; Zhang Ling; Yu Gang; Zhou Liang; Tan Jiansong

2010-01-01

Recently a new method for simulating the thermal loading on pistons of diesel engines was reported. The spatially shaped high power laser is employed as the heat source, and some preliminary experimental and numerical work was carried out. In this paper, a further effort was made to extend this simulation method to some other important engine parts such as cylinder heads. The incident Gaussian beam was transformed into concentric multi-circular patterns of specific intensity distributions, with the aid of diffractive optical elements (DOEs). By incorporating the appropriate repetitive laser pulses, the designed transient temperature fields and thermal loadings in the engine parts could be simulated. Thermal-structural numerical models for pistons and cylinder heads were built to predict the transient temperature and thermal stress. The models were also employed to find the optimal intensity distributions of the transformed laser beam that could produce the target transient temperature fields. Comparison of experimental and numerical results demonstrated that this systematic approach is effective in simulating the thermal loading on the engine parts.

Analytical model for calculation of the thermo hydraulic parameters in a fuel rod assembly

Energy Technology Data Exchange (ETDEWEB)

Cesna, B., E-mail: benas@mail.lei.l [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos g. 3, LT-44403 Kaunas (Lithuania)

2010-11-15

Research highlights: {yields} Proposed calculation model can be used for rapid calculation of the bundles with rods spaced by wire wrapping or honey type spacer grids. {yields} Model estimate three flow cross mixture mechanisms. {yields} Program DARS is enable to analyses experimental results. - Abstract: The paper presents the procedure of the cellular calculation of thermo hydraulic parameters of a single-phase gas flow in a fuel rod assembly. The procedure is implemented in the DARS program. The program is intended for calculation of the distribution of the gaseous coolant parameters and wall temperatures in case of arbitrary, geometrically specified, arrangement of the rods in fuel assembly and in case of arbitrary, functionally specified in space, heat release in the rods. In mathematical model the flow cross-section of the channel of intricate shape is conventionally divided to elementary cells formed by straight lines, which connect the centers of rods. Within the limits of a single cell the coolant parameters and the temperature of the corresponding part of the rod surface are assumed constant. The entire fuel assembly is viewed as a system of parallel interconnected channels. Program DARS is illustrated by calculation of a temperature mode of 85-rod assembly with spacers of wire wrapping on the rods.

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 5 discusses the following topics: Lower Cutting System Test Results and Analysis Report; NFBC Loading System Test Results and Analysis Report; Robotic Bridge Transporter Test Results and Analysis Report; RM-10A Remotec Manipulator Test Results and Analysis Report; and Manipulator Transporter Test Results and Analysis Report

Control rod selecting and driving device

International Nuclear Information System (INIS)

Isobe, Hideo.

1981-01-01

Purpose: To simultaneously drive a predetermined number of control rods in a predetermined mode by the control of addresses for predetermined number of control rods and read or write of driving codified data to and from the memory by way of a memory controller. Constitution: The system comprises a control rod information selection device for selecting predetermined control rods from a plurality of control rods disposed in a reactor and outputting information for driving them in a predetermined mode, a control rod information output device for codifying the information outputted from the above device and outputting the addresses to the predetermined control rods and driving mode coded data, and a driving device for driving said predetermined control rods in a predetermined mode in accordance with the codified data outputted from the above device, said control rod infromation output device comprising a memory device capable of storing a predetermined number of the codified data and a memory control device for storing the predetermined number of data into the above memory device at a predetermined timing while successively outputting the thus stored predetermined number of data at a predetermined timing. (Seki, T.)

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

Science.gov (United States)

Oakes, L.C.; Walker, C.S.

1959-12-15

ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

High repetition ration solid state switched CO2 TEA laser employed in industrial ultrasonic testing of aircraft parts

Science.gov (United States)

von Bergmann, Hubertus; Morkel, Francois; Stehmann, Timo

2015-02-01

Laser Ultrasonic Testing (UT) is an important technique for the non-destructive inspection of composite parts in the aerospace industry. In laser UT a high power, short pulse probe laser is scanned across the material surface, generating ultrasound waves which can be detected by a second low power laser system and are used to draw a defect map of the part. We report on the design and testing of a transversely excited atmospheric pressure (TEA) CO2 laser system specifically optimised for laser UT. The laser is excited by a novel solid-state switched pulsing system and utilises either spark or corona preionisation. It provides short output pulses of less than 100 ns at repetition rates of up to 1 kHz, optimised for efficient ultrasonic wave generation. The system has been designed for highly reliable operation under industrial conditions and a long term test with total pulse counts in excess of 5 billion laser pulses is reported.

Orthogonal cutting of laser beam melted parts

Science.gov (United States)

Götze, Elisa; Zanger, Frederik; Schulze, Volker

2018-05-01

The finishing process of parts manufactured by laser beam melting is of high concern due to the lack of surface accuracy. Therefore, the focus of this work lies on the influence of the build-up direction of the parts and their effect on the finishing process. The orthogonal cutting reveals findings in the fields of chip formation, involved forces and temperatures appearing during machining. In the investigations, the cutting depth was varied between 0.05 and 0.15 mm representing a finishing process and the cutting velocity ranges from 30 to 200 m/min depending on the material. The experiments contain the materials stainless steel (AISI 316L), titanium (Ti6Al4V) and nickel-base alloy (IN718). The two materials named latter are of high interest in the aerospace sector and at the same time titanium is used in the medical field due to its biocompatibility. For the materials IN718 and Ti6Al4V a negative rake angle of -7.5° and for stainless steel a rake angle of 12.5° are chosen for the cutting experiments. The results provide the base for processing strategies. Therefore, the specimens were solely laser beam melted without post-processing like heat treatment. The evaluation of the experiments shows that an increase in cutting speed has different effects depending on the material. For stainless steel the measured forces regarding the machining direction to the layers approach the same values. In contrast, the influence of the layers regarding the forces appearing during orthogonal cutting of the materials IN718 and Ti6Al4V differ for lower cutting speeds.

RODDRP - A FORTRAN program for use in control rod calibration by the rod drop method

International Nuclear Information System (INIS)

Wilson, W.E.

1972-01-01

The different methods to measure reactivity which are applicable to control rod calibration are discussed. They include: 1) the positive period method, 2) the rod drop method, 3) the source-jerk method, 4) the rod oscillation method, and 5) the pulsed neutron method. The instrument setup used at WSU for rod drop measurements is presented. To speed up the analysis of power fall-off trace, a FORTRAN IV program called RODDRP was written to simultaneously solve the in-hour equation and relative neutron flux. The procedure for calculating the worth of the rod that produced the power trace is given. The reactivity for each time relative flux point is obtained. Conclusions about the status of the equipment are made

Design, development and characterization studies of a large aperture high power Nd : glass rod amplifier stage

International Nuclear Information System (INIS)

Gopi, N.; Bapna, R.C.; Murali, C.G.; Narayan, B.S.; Dhareshwar, L.J.

1992-01-01

Laser-plasma interaction studies and experiments related to laser driven shocks as well as laser induced inertially confined thermonuclear fusion have resulted in an ever increasing demand for high brightness lasers capable of producing nanosecond pulse with energy of hundreds of kilo joules. High power Nd-glass laser chains with a master oscillator followed by a number of amplifier stages made up of rods, disks, slabs etc. are in operation in many leading laboratories in the world. This report describes the design, development and characterisation studies of a large aperture Nd:glass laser amplifier which is to be incorporated on line with the existing 40 J, 5 ns high power laser chain built for laser-plasma interaction and laser driven shock wave studies in the Laser and Plasma Technology Division. The development work described in this report discusses the design based on optimum material selection, optimisation of various sub components, ease of maintenance and smooth operation. The necessary operational electronics has also been described. The characterization studies mainly include measurement of spatial gain uniformity, thermally induced depolarization effects, and thermal relaxation studies. (author). 37 refs., 14 figs., 5 tabs

Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

International Nuclear Information System (INIS)

Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

1995-01-01

The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the 'low-temperature' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

Energy Technology Data Exchange (ETDEWEB)

Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

1995-08-01

The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B{sub 4}C absorber rod and the stainless steel grid spacers on the `low-temperature` bundle damage initiation and progression. The B{sub 4}C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

Experimental investigations of single and two-phase flow in a heated rod bundle

International Nuclear Information System (INIS)

Barthel, Frank; Franz, Ronald; Hampel, Uwe; Technische Univ. Dresden

2013-01-01

An experimental facility for the study of boiling flows in a 3 x 3 rod bundle geometry was setup. The bundle resembles in essential geometrical parts the geometry in a pressurized water reactor fuel element. The facility is operated with a refrigerant fluid. Beside standard instrumentation for temperature, pressure and flow rate we employed particle image velocimetry for single phase flow studies, gamma ray densitometry for integral gas fraction measurement sand ultrafast X-ray tomography for the study of the void dynamics in the cross-section. Moreover extensive thermo-instrumentation allows axial rod surface temperature measurements for the central heated rod. First results will be discussed in this article. (orig.)

Simulation of vibration modes of the fuel rod damaged due to the grid-to-rod fretting wear

International Nuclear Information System (INIS)

Kim, Kyu Tae; Kim, Kyeong Koo; Jang, Young Ki; Lee, Kyou Seok

1997-01-01

The flow-induced fuel fretting wear observed in some PWRs mainly proceeds in the grid-to-rod contact positions. The grid-to-rod fretting wear in the PWR fuel assembly depends on grid-to-rod gap size, its axial profile and flow-induced vibration. This paper describes the GRIDFORCE program which generates the axially dependent grid-to-rod gap size as a function of burnup. The axially dependent grid-to-rod gap profiles are employed to predict the fuel rod vibration mode shapes by the ANSYS code. With the help of the Paidousis empirical formula, this paper also calculates the fuel rod vibration amplitudes under various supporting conditions, which indicates that the increase of the number of unsupported mid-grids will increase the fuel rod vibration amplitude. On the other hand, the comparison of the predicted vibration mode shapes and the observed mid-grid fretting wear pattern indicates that the 1st and 6th vibration mode shapes under the supporting inactive condition at the mid-grids can simulate the observed mid-grid fretting wear profile. This paper also proposes design guidelines against the grid-to-rod fretting wear. (author). 3 refs., 8 figs

Automated nuclear fuel rod pattern loading system

International Nuclear Information System (INIS)

Lambert, D.V.; Nylund, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

1991-01-01

This patent describes a method for loading fuel rods in a desired pattern. It comprises providing a supply of fuel rods of known enrichments; providing a magazine defining a matrix of elongated slots open at their forward ends for receiving fuel rods; defining a fuel rod feed path; receiving successively one at a time along the feed path fuel rods selected from the supply thereof; verifying successively one at a time along the feed path the identity of the selected fuel rods, the verifying including blocking passage of each selected fuel rod along the feed path until the identity of each selected fuel rod is confirmed as correct; feeding to the magazine successively one at a time along the feed path the selective and verified fuel rods; and supporting and moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

Control rod drives

International Nuclear Information System (INIS)

Ikakura, Hiroaki.

1986-01-01

Purpose: To enable to direct disconnection of control rods upon abnormal temperature rise in the reactor thereby improve the reliability for the disconnecting operation in control rod drives for FBR type reactors upon emergency. Constitution: A diaphragm is disposed to the upper opening of a sealing vessel inserted to the hollow portion of an electromagnet and a rod is secured to the central position of the upper surface. A spring contacts are attached by way of an insulator to the inner surface at the lower portion of an extension pipe and connected with cables for supplying electric power sources respectively to a magnet. If the temperature in the reactor abnormally rises, liquid metals in the sealing vessel are expanded tending to extend the bellows downwardly. However, since they are attracted by the electromagnet, the thermal expansion of the liquid metals exert on the diaphragm prior to the bellows. Thus, the switch between the spring contacts is made open to attain the deenergized state to thereby disconnect the control rod and shutdown the neclear reactor. (Horiuchi, T.)

Grey Rod Test in HANARO Reactor

Energy Technology Data Exchange (ETDEWEB)

Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

2008-08-15

Westinghouse/KAERI/KNF agreed to perform an irradiation test in the HANARO reactor to obtain irradiation data on the new grey rods that will be part of an AP1000 system. As a preliminary test, two samples containing pure Ag (Reference) and Ag-In-Cd materials provided by Westinghouse Electric Company (WEC) were inserted in a KNF irradiation capsule of 07M-13N. The specimens were irradiated for 95.19days (4 cycles) in the CT test hole of the HANARO of a 30MW thermal output to have a fast neutron fluence of 1.11x10{sup 21}(n/cm{sup 2}) (E>1.0MeV). This report provides all the test conditions and data obtained during the irradiation test of the grey rods in HANARO requested by Westinghouse. The test was prepared according to the meeting minutes (June 26, 2007) and the on-going subject test was stopped midway by the request of Westinghouse.

Method for compacting spent nuclear reactor fuel rods

International Nuclear Information System (INIS)

Wachter, W.J.

1988-01-01

In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

Stable TEM00-mode Nd:YAG solar laser operation by a twisted fused silica light-guide

Science.gov (United States)

Bouadjemine, R.; Liang, D.; Almeida, J.; Mehellou, S.; Vistas, C. R.; Kellou, A.; Guillot, E.

2017-12-01

To improve the output beam stability of a TEM00-mode solar-pumped laser, a twisted fused silica light-guide was used to achieve uniform pumping along a 3 mm diameter and 50 mm length Nd:YAG rod. The concentrated solar power at the focal spot of a primary parabolic mirror with 1.18 m2 effective collection area was efficiently coupled to the entrance aperture of a 2D-CPC/2V-shaped pump cavity, within which the thin laser rod was pumped. Optimum solar laser design parameters were found through ZEMAX© non-sequential ray-tracing and LASCAD© laser cavity analysis codes. 2.3 W continuous-wave TEM00-mode 1064 nm laser power was measured, corresponding to 1.96 W/m2 collection efficiency and 2.2 W laser beam brightness figure of merit. Excellent TEM00-mode laser beam profile at M2 ≤ 1.05 and very good output power stability of less than 1.6% were achieved. Heliostat orientation error dependent laser power variation was considerably less than previous solar laser pumping schemes.

Control rod drives

International Nuclear Information System (INIS)

Asano, Hiromitsu.

1979-01-01

Purpose: To drive control rods at an optimum safety speed corresponding to the reactor core output. Constitution: The reactor power is detected by a neutron detector and the output signal is applied to a process computer. The process computer issues a signal representing the reactor core output, which is converted through a function generator into a signal representing the safety speed of control rods. The converted signal is further supplied to a V/F converter and converted into a pulse signal. The pulse signal is inputted to a step motor driving circuit, which actuates a step motor to operate the control rods always at a safety speed corresponding to the reactor core power. (Furukawa, Y.)

Individual nuclear fuel rod weighing system

International Nuclear Information System (INIS)

Fogg, J. L.; Howell, C. A.; Smith, J. H.; Vining, G. E.

1985-01-01

An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

Measuring device for control rod driving time

International Nuclear Information System (INIS)

Tanaka, Kazuhiko; Hanabusa, Masatoshi.

1993-01-01

The present invention concerns a measuring device for control driving time having a function capable of measuring a selected control rod driving time and measuring an entire control rod driving time simultaneously. A calculation means and a store means for the selected rod control rod driving time, and a calculation means and a store means for the entire control rod driving time are disposed individually. Each of them measures the driving time and stores the data independent of each other based on a selected control rod insert ion signal and an entire control rod insertion signal. Even if insertion of selected and entire control rods overlaps, each of the control rod driving times can be measured reliably to provide an advantageous effect capable of more accurately conducting safety evaluation for the nuclear reactor based on the result of the measurement. (N.H.)

Growth and Morphology of Rod Eutectics

Energy Technology Data Exchange (ETDEWEB)

Jing Teng; Shan Liu; R. Trivedi

2008-03-17

The formation of rod eutectic microstructure is investigated systematically in a succinonitrile-camphor alloy of eutectic composition by using the directional solidification technique. A new rod eutectic configuration is observed in which the rods form with elliptical cylindrical shape. Two different orientations of the ellipse are observed that differ by a 90{sup o} rotation such that the major and the minor axes are interchanged. Critical experiments in thin samples, where a single layer of rods forms, show that the spacing and orientation of the elliptic rods are governed by the growth rate and the sample thickness. In thicker samples, multi layers of rods form with circular cross-section and the scaling law between the spacing and velocity predicted by the Jackson and Hunt model is validated. A theoretical model is developed for a two-dimensional array of elliptical rods that are arranged in a hexagonal or a square array, and the results are shown to be consistent with the experimental observations. The model of elliptic rods is also shown to reduce to that for the circular rod eutectic when the lengths of the two axes are equal, and to the lamellar eutectic model when one of the axes is much larger than the other one.

Triggering and guiding high-voltage large-scale leader discharges with sub-joule ultrashort laser pulses

International Nuclear Information System (INIS)

Pepin, H.; Comtois, D.; Vidal, F.; Chien, C.Y.; Desparois, A.; Johnston, T.W.; Kieffer, J.C.; La Fontaine, B.; Martin, F.; Rizk, F.A.M.; Potvin, C.; Couture, P.; Mercure, H.P.; Bondiou-Clergerie, A.; Lalande, P.; Gallimberti, I.

2001-01-01

The triggering and guiding of leader discharges using a plasma channel created by a sub-joule ultrashort laser pulse have been studied in a megavolt large-scale electrode configuration (3-7 m rod-plane air gap). By focusing the laser close to the positive rod electrode it has been possible, with a 400 mJ pulse, to trigger and guide leaders over distances of 3 m, to lower the leader inception voltage by 50%, and to increase the leader velocity by a factor of 10. The dynamics of the breakdown discharges with and without the laser pulse have been analyzed by means of a streak camera and of electric field and current probes. Numerical simulations have successfully reproduced many of the experimental results obtained with and without the presence of the laser plasma channel

Status of rod consolidation, 1988

International Nuclear Information System (INIS)

Bailey, W.J.

1989-01-01

It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs

Inspecting method for fuel rods

International Nuclear Information System (INIS)

Watanabe, Masaaki; Kogure, Sumio.

1976-01-01

Purpose: To precisely detect the response of flaw in clad tube and submerged fuel pellets from a relationship between the surface of fuel rod and internal signal. Constitution: Ultrasonic reflected waves from the surface of fuel rods and the interior are detected and either one of fuel rod or ultrasonic flaw detecting contact is rotated to thereby precisely detect the response of the flaw of clad tube and submerged fuel pellets from a relationship between said surface and the interior. It will be noted that the ultrasonic flaw detecting contact used is of the line-focus type, the incident angle of ultrasonic wave from the ultrasonic flaw detecting contact relative to the fuel rod is the angle of skew, that is, the ultrasonic flaw detecting contact is not perpendicular to a center axis of the fuel rod but is slightly displace. That is, the use of the aforesaid contact may facilitate discrimination between the surface flaw of the fuel rod and the response of submergence, and in addition, the employment of the aforesaid incident angle makes it hard to receive reflected waves from the surface of the fuel rod which is great in terms of energy to facilitate discrimination of waves responsive to submergence. (Kawakami, Y.)

Control rod position detection device

International Nuclear Information System (INIS)

Akita, Haruo; Ogiwara, Sakae.

1996-01-01

The device of the present invention is used in a back-up shut down system of an LMFBR type reactor which is easy for maintenance, has high reliability and can recognize the position of control rods accurately. Namely, a permanent magnet is disposed to a control rod extension tube connected to the lower portion of the control rod. The detector guide tube is disposed in the vicinity of the control rod extension tube. A detector having a detection coil is inserted into a detector tube. With such constitution, the control rod can be detected at one position using the following method. (1) the movement of the magnetic field of the permanent magnet is detected by the detection coil. (2) a plurality of grooves are formed on the control rod extension tube, and the movement of the grooves is detected. In addition, the detection coil is inserted into the detector guide tube, and the signals from the detection coil are inputted to a signal processing circuit disposed at the outside of the reactor vessel using an MI cable to enable the maintenance of the detector. Further, if the detector comprises a detection coil and an excitation coil, the position of a dropped control rod can be recognized at a plurality of points. (I.S.)

Rope wind-up type control rod

International Nuclear Information System (INIS)

Tsuji, Teruaki; Watanabe, Shigeru.

1979-01-01

Purpose: To hold a control rod at a certain position even if the sealed cover of the rod drive mechanism should fail. Constitution: A plurality of friction plates, engaging wheels and a threaded shaft are provided to the wind-up drum for winding up a rope which moves the control rod up and down. While the control rod is adapted to drop by its own weight upon insertion, it is adapted to stop at a predetermined position exactly with no shocks by gradually increasing braking force by the sliding friction caused from the friction plates or the like. A ratch mechanism is provided to the upper portion of the control rod so that the top of the ratch piece may automatically engage the guide passage wall of the control rod upon uncontrolled running of the control rod to prevent further uncontrolled running thereof. (Ikeda, J.)

Development of solar concentrators for high-power solar-pumped lasers.

Science.gov (United States)

Dinh, T H; Ohkubo, T; Yabe, T

2014-04-20

We have developed unique solar concentrators for solar-pumped solid-state lasers to improve both efficiency and laser output power. Natural sunlight is collected by a primary concentrator which is a 2  m×2  m Fresnel lens, and confined by a cone-shaped hybrid concentrator. Such solar power is coupled to a laser rod by a cylinder with coolant surrounding it that is called a liquid light-guide lens (LLGL). Performance of the cylindrical LLGL has been characterized analytically and experimentally. Since a 14 mm diameter LLGL generates efficient and uniform pumping along a Nd:YAG rod that is 6 mm in diameter and 100 mm in length, 120 W cw laser output is achieved with beam quality factor M2 of 137 and overall slope efficiency of 4.3%. The collection efficiency is 30.0  W/m2, which is 1.5 times larger than the previous record. The overall conversion efficiency is more than 3.2%, which can be comparable to a commercial lamp-pumped solid-state laser. The concept of the light-guide lens can be applied for concentrator photovoltaics or other solar energy optics.

ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

Science.gov (United States)

Young, J.N.

1958-04-22

An electromagnetic apparatus for moving a rod-like member in small steps in either direction is described. The invention has particular application in the reactor field where the reactor control rods must be moved only a small distance and where the use of mechanical couplings is impractical due to the high- pressure seals required. A neutron-absorbing rod is mounted in a housing with gripping uaits that engage the rod, and coils for magnetizing the gripping units to make them grip, shift, and release the rod are located outside the housing.

Individual nuclear fuel rod weighing system

International Nuclear Information System (INIS)

Fogg, J.L.; Smith, J.H.; Vining, G.E.; Howell, C.A.

1985-01-01

An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor is discussed. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

Control rods in LMFBRs: a physics assessment

International Nuclear Information System (INIS)

McFarlane, H.F.; Collins, P.J.

1982-08-01

This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B 4 C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined

Evaluation of the Westinghouse 10B depletion for BWR control rods

International Nuclear Information System (INIS)

Vallgren, Christina

2008-03-01

The aim of this work was to establish the 10 B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10 B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10 B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10 B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10 B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10 B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10 B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10 B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10 B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10 B depletion models on the process computers showed a difference of about 20 %. By using the new 10 B depletion model m POLCA7, it was found that the 10 B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B 4 C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight

Evaluation of the Westinghouse 10B depletion for BWR control rods

Energy Technology Data Exchange (ETDEWEB)

Vallgren, Christina

2008-03-15

The aim of this work was to establish the 10B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10B depletion models on the process computers showed a difference of about 20 %. By using the new 10B depletion model m POLCA7, it was found that the 10B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B{sub 4}C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight difference in

Investigation into the influence of laser energy input on selective laser melted thin-walled parts by response surface method

Science.gov (United States)

Liu, Yang; Zhang, Jian; Pang, Zhicong; Wu, Weihui

2018-04-01

Selective laser melting (SLM) provides a feasible way for manufacturing of complex thin-walled parts directly, however, the energy input during SLM process, namely derived from the laser power, scanning speed, layer thickness and scanning space, etc. has great influence on the thin wall's qualities. The aim of this work is to relate the thin wall's parameters (responses), namely track width, surface roughness and hardness to the process parameters considered in this research (laser power, scanning speed and layer thickness) and to find out the optimal manufacturing conditions. Design of experiment (DoE) was used by implementing composite central design to achieve better manufacturing qualities. Mathematical models derived from the statistical analysis were used to establish the relationships between the process parameters and the responses. Also, the effects of process parameters on each response were determined. Then, a numerical optimization was performed to find out the optimal process set at which the quality features are at their desired values. Based on this study, the relationship between process parameters and SLMed thin-walled structure was revealed and thus, the corresponding optimal process parameters can be used to manufactured thin-walled parts with high quality.

Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M; Hernborg, G; Flinta, J E

1964-08-15

The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within {+-} 5 per cent.

Development of a control rod drive

International Nuclear Information System (INIS)

1991-01-01

In the period under review, the computer codes required for transients calculation have been completed, as well as the programs for modelling and testing the hot-gas temperature control by means of combined core rod and reflector rod operation. The specification of requirements to be fulfilled by the rod drive computer and the neutron flux measuring system has been done relying essentially on the data obtained by the transients calculations performed and the resulting informations on operating conditions. The work for optimization of the core rod drive with regard to rod driving speeds and the 'three-point switch' with hysteresis for controlled, automatic core rod operation has been concentrating on the case of specified, normal operation of the reactor. (orig./DG) [de

Self-Assembly of Rod-Coil Block Copolymers

National Research Council Canada - National Science Library

Jenekhe, S

1999-01-01

... the self-assembly of new rod-coil diblock, rod- coil-rod triblock, and coil-rod-coil triblock copolymers from solution and the resulting discrete and periodic mesostmctares with sizes in the 100...

Ejected control rod and rods drop measurements during Mochovce startup physical tests

International Nuclear Information System (INIS)

Minarcin, Miroslav; Elko, Marek

1998-01-01

Paper deals with measurements of asymmetric reactivity insertion into the reactor core that were carried out during physical startup tests of Mochovce Unit 1 in June 1998. Control rods worth measurements with one and two rods s tucked in upper limit and worth measurement of one control rod from group 6 'ejected' from the reactor core are discussed. During the experiments neutron flux was measured by four ionisation chambers (three of them were placed symmetrically around the reactor core). Results of measurements and influence of asymmetric reactivity influence on ionisation chambers response are presented in the paper. (Authors)

Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

International Nuclear Information System (INIS)

Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland; Helmut Kuhl

2015-01-01

While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs

Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

Energy Technology Data Exchange (ETDEWEB)

Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

2015-05-15

While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

Numerical simulation of complex part manufactured by selective laser melting process

Science.gov (United States)

Van Belle, Laurent

2017-10-01

Selective Laser Melting (SLM) process belonging to the family of the Additive Manufacturing (AM) technologies, enable to build parts layer by layer, from metallic powder and a CAD model. Physical phenomena that occur in the process have the same issues as conventional welding. Thermal gradients generate significant residual stresses and distortions in the parts. Moreover, the large and complex parts to manufacturing, accentuate the undesirable effects. Therefore, it is essential for manufacturers to offer a better understanding of the process and to ensure production reliability of parts with high added value. This paper focuses on the simulation of manufacturing turbine by SLM process in order to calculate residual stresses and distortions. Numerical results will be presented.

Monitoring device for withdrawing control rods

International Nuclear Information System (INIS)

Higashigawa, Yuichi.

1985-01-01

Purpose: To improve the sensitivity and the responsivity to an equivalent extent to those in the case where local power range monitors are densely arranged near each of the control rods, with no actual but pseudo increase of the number of local power range monitors. Constitution: The monitor arrangement is patterned by utilizing the symmetricity of the reactor core and stored in a monitor designating device. The symmetricity of control rods to be selected and withdrawn by an operator is judged by a control rod symmetry monitoring device, while the symmetricity of the withdrawn control rods is judged by a control rod withdrawal state monitoring device. Then, only when both of the devices judge the symmetricity, the control rods are subjected to gang driving by the control rod drive mechanisms. In this way, monitoring at a high sensitivity and responsivity is enabled with no increase for the number of monitors. (Yoshino, Y.)

Nondestructive assay of HTGR fuel rods

International Nuclear Information System (INIS)

Menlove, H.O.

1974-01-01

Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

Bistable direction switching in an off-axis pumped continuous wave ruby laser

Science.gov (United States)

Afzal, R. Sohrab; Lawandy, N. M.

1988-01-01

A report is presented of the observation of hysteretic bistable direction switching in a single-mode CW ruby laser system. This effect is only observed when the pump beam which is focused into the ruby rod is misaligned with respect to the rod end faces. At low pump powers, the ruby lases in a mode nearly collinear with the pump axis. At a higher pump power the ruby switches to a mode that is collinear with the rod end faces and preserves the original polarization. The effect is large enough to switch the beam by an angle equal to twice the diffraction angle. The observations show that under steady-state pumping, a CW ruby laser can exhibit bistable operation in its output direction and power. A calculation using the heat equation with two concentric cylinders with one as a heat source (pump laser) and the outer wall of the other held at 77 K, gives an increase in core temperature of about 0.01 K. Therefore, the increase in temperature is not large enough to change the index of refraction to account for such large macroscopic effects.

Inspection system for Zircaloy clad fuel rods

International Nuclear Information System (INIS)

Yancey, M.E.; Porter, E.H.; Hansen, H.R.

1975-10-01

A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

Paediatric laser dentistry. Part 1: General introduction.

Science.gov (United States)

Caprioglio, C; Olivi, G; Genovese, M D

2017-03-01

Knowledge of the physical characteristics of different laser lights and optical and thermal properties of oral tissues is very important to understand the interaction of dental lasers with biological tissues. Choosing the correct dental laser is crucial to match specific wavelengths with target chromophores of different tissues; this affinity makes laser irradiation selective and therefore minimally invasive. Various types of lasers are used in dentistry, offering a viable alternative to low and high-speed handpieces and surgical blades, and also minimising fear and discomfort of the patient. Lasers can provide innovative and minimally invasive therapies in different branches of dentistry including preventive and restorative dentistry, traumatic injury treatments and surgical procedures. Laser has also biostimulating and anti-inflammatory effects, as well as analgesic effect.

Stick-Slip of Lightly Loaded Rock. Part 1. Dilatancy and Shearing Behavior of Assemblages of Rods. Part 2

Science.gov (United States)

1975-02-04

particularly at light normal stresses. Such self-induced vibrations were recog- nized in the nineteenth century as the source of music from violin ...of slip zones developing during a test may be calculated from the number of possible combinations. For exam - ple, let us assume that due to end...combinations. For exam - ple, let us assume that due to end restraint only the four central planes in the array of one inch rods art, eligil)le for slip

Numerical simulation of 900 MW control rods impact friction vibration and wear

International Nuclear Information System (INIS)

Jacquart, G.

1993-12-01

Impact-friction vibrations and wear have motivated a great research and development program aiming at understanding the impact and vibration behaviour of these components through experimental and numerical works. This report presents a numerical simulation of the vibrations of a single control rod and of a whole control cluster. Excitation sources for this component are due to hydraulic forces and are situated in the lower part of the rods and in the part of the cluster. Some parametric computations have been carried out on a single rod, to evaluate the effect of the lower excitation source. Different excitation levels, different eccentricities or static forces have been computed and compared to measurements on the MAGALY mock-up representing a complete rod cluster. A numerical model for the complete cluster allowed the evaluation of the upper excitation source effects. This source appears to be less powerful than the lower one. These results have been validated by comparison with MAGALY measurements. At last, some computations were performed with a model of the complete cluster, taking into account the both excitation sources. A parametric study on eccentricity and static forces has been carried out. A comparison with MAGALY measurements seems to be fairly fitting, showing that the numerical results are of the right order of magnitude. Through this numerical study, we have shown that numerical simulation of a complete control rod cluster could be lead, and we have obtained some new informations about impact forces and wear rates that need to be confirmed by more computational or experimental works or in-situ measurements. (author). 10 annexes, 11 refs

A Spectral Mapping Signature for the Rapid Ohia Death (ROD Pathogen in Hawaiian Forests

Directory of Open Access Journals (Sweden)

Gregory P. Asner

2018-03-01

Full Text Available Pathogenic invasions are a major source of change in both agricultural and natural ecosystems. In forests, fungal pathogens can kill habitat-generating plant species such as canopy trees, but methods for remote detection, mapping and monitoring of such outbreaks are poorly developed. Two novel species of the fungal genus Ceratocystis have spread rapidly across humid and mesic forests of Hawaiʻi Island, causing widespread mortality of the keystone endemic canopy tree species, Metrosideros polymorpha (common name: ʻōhiʻa. The process, known as Rapid Ohia Death (ROD, causes browning of canopy leaves in weeks to months following infection by the pathogen. An operational mapping approach is needed to track the spread of the disease. We combined field studies of leaf spectroscopy with laboratory chemical studies and airborne remote sensing to develop a spectral signature for ROD. We found that close to 80% of ROD-infected plants undergo marked decreases in foliar concentrations of chlorophyll, water and non-structural carbohydrates, which collectively result in strong consistent changes in leaf spectral reflectance in the visible (400–700 nm and shortwave-infrared (1300–2500 nm wavelength regions. Leaf-level results were replicated at the canopy level using airborne laser-guided imaging spectroscopy, with quantitative spectral separability of normal green-leaf canopies from suspected ROD-infected brown-leaf canopies in the visible and shortwave-infrared spectrum. Our results provide the spectral–chemical basis for detection, mapping and monitoring of the spread of ROD in native Hawaiian forests.

Seismic scrammability of HTTR control rods

International Nuclear Information System (INIS)

Nishiguchi, I.; Iyoku, T.; Ito, N.; Watanabe, Y.; Araki, T.; Katagiri, S.

1990-01-01

Scrammability tests on HTTR (High-Temperature Engineering Test Reactor) control rods under seismic conditions have been carried out and seismic conditions influences on scram time as well as functional integrity were examined. A control rod drive located in a stand-pipe at the top of a reactor vessel, raises and lowers a pair of control rods by suspension cables. Each flexible control rod consists of 10 neutron absorber sections held together by a metal spine passing through the center. It falls into a hole in graphite blocks due to gravity at scram. In the tests, a full scale control rod drive and a pair of control rods were employed with a column of graphite blocks in which holes for rods were formed. Blocks misalignment and contact with the hole surface during earthquakes were considered as major causes of disturbance in scram time. Therefore, the following parameters were set up in the tests: excitation direction, combination or horizontal and vertical excitation, acceleration, frequency and block to block gaps. Main results obtained from tests are as follow. 1) Every scram time obtained under the design conditions was within 6 seconds. On the contrary, the scram times were 5.2 seconds when there were no vibration. Therefore, it was concluded that the seismic effects on scram time were not significant. 2) Scram time became longer with increase in both acceleration and horizontal excitation frequency, and control rods fell very smoothly without any jerkiness. This suggests that collision between control rods and hole surface is the main disturbing factor of falling motion. 3) Mechanical and functional integrity of control rod drive mechanism, control rods and graphite blocks was confirmed after 140 seismic scrammability tests. (author). 10 figs, 1 tab

Airclad fiber laser technology

DEFF Research Database (Denmark)

Hansen, Kim P.; Olausson, Christina Bjarnal Thulin; Broeng, Jes

2011-01-01

High-power fiber lasers and amplifiers have gained tremendous momentum in the last 5 years. Many of the traditional manufacturers of gas and solid-state lasers are now pursuing the fiber-based systems, which are displacing the conventional technology in many areas. High-power fiber laser systems...... require reliable fibers with large cores, stable mode quality, and good power handling capabilities-requirements that are all met by the airclad fiber technology. In the present paper we go through many of the building blocks needed to build high-power systems and we show an example of a complete airclad...... laser system. We present the latest advancements within airclad fiber technology including a new 100 m single-mode polarization-maintaining rod-type fiber capable of amplifying to megawatt power levels. Furthermore, we describe the novel airclad-based pump combiners and their use in a completely...

Control rod testing apparatus

International Nuclear Information System (INIS)

Gaunt, R.R.; Ashman, C.M.

1987-01-01

A control rod testing apparatus is described comprising: a first guide means having a vertical cylindrical opening for grossly guiding a control rod; a second guide means having a vertical cylindrical opening for grossly guiding a control rod. The first and second guide means are supported at axially spaced locations with the openings coaxial; and a substantially cylindrical subassembly having a vertical cylindrical opening therethrough. The subassembly is trapped coaxial with and between the first and second guide means, and the subassembly radially floats with respect to the first and second guide means

Nuclear reactor fuel rod

International Nuclear Information System (INIS)

Busch, H.; Mindnich, F.R.

1973-01-01

The fuel rod consists of a can with at least one end cap and a plenum spring between this cap and the fuel. To prevent the hazard that a eutectic mixture is formed during welding of the end cap, a thermal insulation is added between the end cap and plenum spring. It consists of a comical extension of the end cap with a terminal disc against which the spring is supported. The end cap, the extension, and the disc may be formed by one or several pieces. If the disc is separated from the other parts it may be manufactured from chrome steel or VA steel. (DG) [de

Development of the automatic control rod operation system for JOYO. Verification of automatic control rod operation guide system

International Nuclear Information System (INIS)

Terakado, Tsuguo; Suzuki, Shinya; Kawai, Masashi; Aoki, Hiroshi; Ohkubo, Toshiyuki

1999-10-01

The automatic control rod operation system was developed to control the JOYO reactor power automatically in all operation modes(critical approach, cooling system heat up, power ascent, power descent), development began in 1989. Prior to applying the system, verification tests of the automatic control rod operation guide system was conducted during 32nd duty cycles of JOYO' from Dec. 1997 to Feb. 1998. The automatic control rod operation guide system consists of the control rod operation guide function and the plant operation guide function. The control rod operation guide function provides information on control rod movement and position, while the plant operation guide function provide guidance for plant operations corresponding to reactor power changes(power ascent or power descent). Control rod insertion or withdrawing are predicted by fuzzy algorithms. (J.P.N.)

Control rod drive shaft latch

International Nuclear Information System (INIS)

Thorp, A.G. II.

1976-01-01

A latch mechanism is operated by differential pressure on a piston to engage the drive shaft for a control rod in a nuclear reactor, thereby preventing the control rod from being ejected from the reactor in case of failure of the control rod drive mechanism housing which is subjected to the internal pressure in the reactor vessel. 6 claims, 4 drawing figures

Thermal and Fluid Mechanical Investigation of an Internally Cooled Piston Rod

Science.gov (United States)

Klotsche, K.; Thomas, C.; Hesse, U.

2017-08-01

The Internal Cooling of Reciprocating Compressor Parts (ICRC) is a promising technology to reduce the temperature of the thermally stressed piston and piston rod of process gas compressors. The underlying heat transport is based on the flow of a two-phase cooling medium that is contained in the hollow reciprocating assembly. The reciprocating motion forces the phases to mix, enabling an enhanced heat transfer. In order to investigate this heat transfer, experimental results from a vertically reciprocating hollow rod are presented that show the influence of different liquid charges for different working temperatures. In addition, pressure sensors are used for a crank angle dependent analysis of the fluid mechanical processes inside the rod. The results serve to investigate the two-phase flow in terms of the velocity and distribution of the liquid and vapour phase for different liquid fractions.

Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

International Nuclear Information System (INIS)

Becker, Kurt M.; Hernborg, G.; Flinta, J.E.

1964-08-01

The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within ± 5 per cent

Application of pattern recognition techniques to the detection of the Phenix reactor control rods vibrations

International Nuclear Information System (INIS)

Zwingelstein, G.; Deat, M.; Le Guillou, G.

1979-01-01

The incipient detection of control rods vibrations is very important for the safety of the operating plants. This detection can be achieved by an analysis of the peaks of the power spectrum density of the neutron noise. Pattern Recognition techniques were applied to detect the rod vibrations which occured at the fast breeder Phenix (250MWe). In the first part we give a description of the basic pattern which is used to characterize the behavior of the plant. The pattern is considered as column vector in n dimensional Euclidian space where the components are the samples of the power spectral density of the neutron noise. In the second part, a recursive learning procedure of the normal patterns which provides the mean and the variance of the estimates is described. In the third part the classification problem has been framed in terms of a partitioning procedure in n dimensional space which encloses regions corresponding to normal operations. This pattern recognition scheme was applied to the detection of rod vibrations with neutron data collected at the Phenix site before and after occurence of the vibrations. The analysis was carried out with a 42-dimensional measurement space. The learned pattern was estimated with 150 measurement vectors which correspond to the period without vibrations. The efficiency of the surveillance scheme is then demonstrated by processing separately 119 measurement vectors recorded during the rod vibration period

Volumetric imaging of rod and cone photoreceptor structure with a combined adaptive optics-optical coherence tomography-scanning laser ophthalmoscope

Science.gov (United States)

Wells-Gray, Elaine M.; Choi, Stacey S.; Zawadzki, Robert J.; Finn, Susanna C.; Greiner, Cherry; Werner, John S.; Doble, Nathan

2018-03-01

We have designed and implemented a dual-mode adaptive optics (AO) imaging system that combines spectral domain optical coherence tomography (OCT) and scanning laser ophthalmoscopy (SLO) for in vivo imaging of the human retina. The system simultaneously acquires SLO frames and OCT B-scans at 60 Hz with an OCT volume acquisition time of 4.2 s. Transverse eye motion measured from the SLO is used to register the OCT B-scans to generate three-dimensional (3-D) volumes. Key optical design considerations include: minimizing system aberrations through the use of off-axis relay telescopes, conjugate pupil plane requirements, and the use of dichroic beam splitters to separate and recombine the OCT and SLO beams around the nonshared horizontal scanning mirrors. To demonstrate system performance, AO-OCT-SLO images and measurements are taken from three normal human subjects ranging in retinal eccentricity from the fovea out to 15-deg temporal and 20-deg superior. Also presented are en face OCT projections generated from the registered 3-D volumes. The ability to acquire high-resolution 3-D images of the human retina in the midperiphery and beyond has clinical importance in diseases, such as retinitis pigmentosa and cone-rod dystrophy.

Nondestructive post-irradiation examination of Loop-1, S1 and B1 rods

International Nuclear Information System (INIS)

Bratton, R.L.

1997-05-01

As a part of the Pacific Northwest National Laboratory's Tritium Target Development Program, eleven tritium target rods were irradiated in the Advanced Test Reactor located at the Idaho National Engineering and Environmental Laboratory during 1991. Both nondestructive and destructive post-irradiation examination on all eleven rods was planned under the Tritium Target Development Program. Funding for the program was reduced in 1991 resulting in the early removal of the program experiments before reaching their irradiation goals. Post-irradiation examination was only performed on one of the irradiated rods at the Pacific Northwest National Laboratory before the program was terminated in 1992. On December 6, 1995, the Secretary of Energy announced the pursuit of the Commercial Light-Water Reactor option for producing tritium establishing the Tritium Target Qualification Program at the Pacific Northwest National Laboratory. This program decided to pursue nondestructive and destructive post-irradiation examination of the ten remaining rods from the previous program. The ten rods comprise three experiments. The Loop-1 experiment irradiated eight target rods in a loop configuration for 217 irradiation days. The other two rods were irradiated in two separate irradiation experiments, designated as S1 and B1 for 143 effective full-power days, but at different power levels. After the ten rods were transferred from the ATR Canal to the Hot Fuels Examination Facility, the following examinations were performed: (1) visual examination and photography; (2) neutron radiography; (3) axial gamma scanning; (4) contact profilometry measurement; (5) bow and length measurements; (6) rod puncture and plenum gas analysis/measurement of plenum gas quantity; (7) void volume determination; and (8) internal pressure determination. This report presents the data collected during these examinations

Reconstitutable control rod spider assembly

International Nuclear Information System (INIS)

Shallenberger, J.M.; Ferian, S.J.

1990-01-01

A reconstitutable control rod/spider assembly includes a hollow connecting finger of the spider having a pair of opposing flat segments formed on the interior thereof and engaging a pair of opposing flat sectors formed on the exterior of a stem extending form the upper end of control rod. The stem also has an externally-threaded portion engaging a nut and a pilot aligning portion for the nut. The nut has a radially flexible and expandable thread-defining element captured in its bore. The segments and sectors allow the rod to be removed and reattached after turning through 180 0 to allow more even wear on the rod. (author)

Estimation of irradiated control rod worth

International Nuclear Information System (INIS)

Varvayanni, M.; Catsaros, N.; Antonopoulos-Domis, M.

2009-01-01

When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber's burnup. The suggested methodology is based on measurements of the rod's worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.

The study of the laser parameters and environment variables effect on mechanical properties of high compact parts elaborated by selective laser melting 316L powder

International Nuclear Information System (INIS)

Zhang, Baicheng; Dembinski, Lucas; Coddet, Christian

2013-01-01

In this work, a systematic analysis of the main parameters for the selective laser melting (SLM) of a commercial stainless steel 316L powder was conducted to improve the mechanical properties and dimensional accuracy of the fabricated parts. First, the effects of the processing parameters, such as the laser beam scanning velocity, laser power, substrate condition and thickness of the powder layer, on the formation of single tracks for achieving a continuous melting and densification of the material were analysed. Then, the influence of the environmental conditions (gas nature) and of the preheating temperature on the density and dimensional accuracy of the parts was considered. The microstructural features of the SLM SS 316L parts were carefully observed to elucidate the melting-solidification mechanism and the thermal history, which are the basis of the manufacturing process. Finally, the mechanical properties of the corresponding material were also determined

Rod cluster having improved vane configuration

International Nuclear Information System (INIS)

Shockling, L.A.; Francis, T.A.

1989-01-01

This patent describes a pressurized water reactor vessel, the vessel defining a predetermined axial direction of the flow of coolant therewithin and having plural spider assemblies supporting, for vertical movement within the vessel, respective clusters of rods in spaced, parallel axial relationship, parallel to the predetermined axial direction of coolant flow, and a rod guide for each spider assembly and respective cluster of rods. The rod guide having horizontally oriented support plates therewithin, each plate having an interior opening for accommodating axial movement therethrough of the spider assembly and respective cluster of rods. The opening defining plural radially extending channels and corresponding parallel interior wall surfaces of the support plate

MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

International Nuclear Information System (INIS)

Hernandez L, H.; Ortiz V, J.

2003-01-01

In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

Sucker rod motor

Energy Technology Data Exchange (ETDEWEB)

Radzalov, N N; Radzhabov, N A

1983-01-01

The motor consists of rollers mounted on the wellmouth and connected by a flexible rink. Reciprocating mechanism is in the form of a horizontal non-mobile single-side operation cylinder, inside which a plunger and rod are mounted. The working housing of the hydrocylinder is connected to a gas-hydr aulic batter, and when running is connected via plunger to the high pressure source; running in reverse it is connected with a safety valve and automatic control unit. The unit is equipped with a reducer and a mechanical transformer consisting of screw and nut, and which is shutoff with a single-side lining. The plunger rod consists of an auger-like unit. The high pressure source is provided by the injection line of the sucker rod that has been equipped with a reverse valve.

Duke Power Company's control rod wear program

International Nuclear Information System (INIS)

Culp, D.C.; Kitlan, M.S. Jr.

1990-01-01

Recent examinations performed at several foreign and domestic pressurized water reactors have identified significant control rod cladding wear, leading to the conclusion that previously believed control rod lifetimes are not attainable. To monitor control rod performance and reduce safety concerns associated with wear, Duke Power Company has developed a comprehensive control rod wear program for Ag-In-Cd and boron carbide (B 4 C) rods at the McGuire and Catawba nuclear stations. Duke Power currently uses the Westinghouse 17 x 17 Ag-In-Cd control rod design at McGuire Unit 1 and the Westinghouse 17 x 17 hybrid B 4 C control rod design with a Ag-In-Cd tip at McGuire Unit 2 and Catawba Units 1 and 2. The designs are similar, with the exception of the absorber material and clad thickness. There are 53 control rods per unit

Fuel rod pellet loading head

International Nuclear Information System (INIS)

Howell, T.E.

1975-01-01

An assembly for loading nuclear fuel pellets into a fuel rod comprising a loading head for feeding pellets into the open end of the rod is described. The pellets rest in a perforated substantially V-shaped seat through which air may be drawn for removal of chips and dust. The rod is held in place in an adjustable notched locator which permits alignment with the pellets

Non-Destructive Measurement of Residual Strain in Connecting Rods Using Neutrons

Energy Technology Data Exchange (ETDEWEB)

Ikeda, Tomohiro [Honda R& D; Bunn, Jeffrey R. [ORNL; Fancher, Christopher M. [ORNL; Seid, Alan [Honda R& D; Motani, Ryuta [Honda R& D; Matsuda, Hideki [Honda R& D; Okayama, Tatsuya [Honda R& D

2018-04-01

Increasing the strength of materials is effective in reducing weight and boosting structural part performance, but there are cases in where the residual strain generated during the process of manufacturing of high-strength materials results in a decline of durability. It is therefore important to understand how the residual strain in a manufactured component changes due to processing conditions. In the case of a connecting rod, because the strain load on the connecting rod rib sections is high, it is necessary to clearly understand the distribution of strain in the ribs. However, because residual strain is generally measured by using X-ray diffractometers or strain gauges, measurements are limited to the surface layer of the parts. Neutron beams, however, have a higher penetration depth than X-rays, allowing for strain measurement in the bulk material. The research discussed within this paper consists of non-destructive residual strain measurements in the interior of connecting rods using the 2nd Generation Neutron Residual Stress Mapping Facility (NRSF2) at Oak Ridge National Laboratory, measuring the Fe (211) diffraction peak position of the ferrite phase. The interior strain distribution of connecting rod, which prepared under different manufacturing processes, was revealed. By the visualization of interior strains, clear understandings of differences in various processing conditions were obtained. In addition, it is known that the peak width, which is also obtained during measurement, is suggestive of the size of crystallites in the structure; however the peak width can additionally be caused by microstresses and material dislocations.

Rod displacement measurements by x-ray CT and its impact on thermal-hydraulics in tight-lattice rod bundle (Joint research)

International Nuclear Information System (INIS)

Mitsutake, Toru; Misawa, Takeharu; Kureta, Masatoshi; Akimoto, Hajime

2005-06-01

In tight-lattice simulated rod bundles with about 1 mm gap between rods, a rod displacement might affect thermal-hydraulic characteristics since the displacement has a strong impact on the flow area change along the heated section. It should be important to estimate how large the rod position displacement could quantitatively affect critical power for the tight-lattice rod bundle from the point of improvement of prediction capability of subchannel analysis. In the present study, the inside-structure observation of the simulated seven-rod bundle of Reduced Moderation Water Reactor (RMWR) was made through the whole length of the test assembly. Based on the measured rod position data, the relation between the rod position displacement and the heat transfer characteristics was investigated experimentally and through the two kinds of subchannel analysis, the nominal rod position case and the measured rod position case, the effect on the predicted critical power was estimated. The high-energy X-ray computer tomograph (CT) of Fuels Monitoring Facilities (FMF) at the O-arai Engineering Center in Japan Nuclear Cycle Institute (JNC) was applied for the inside-structure observation of the test assembly. The CT view of the cross sections within the test assembly assured the hexagonal rod position arrangement was almost the same as expected by design. The measured data with the X-ray CT facility showed that all rod displacements were small, 0.5 millimeters at maximum and 0.2 millimeters in average. In the heat transfer experiments for the seven-rod bundle, the boiling transition (BT) position and the rod surface temperature behavior was measured. All thermocouples on the center rod downstream from the BT-onset axial height showed almost simultaneous temperature increase due to BT. And the thermocouples located on the same axial heights showed quite similar time-variation behaviors in the vapor cooling heat transfer regime. These results demonstrated the effect of the

Postirradiation examination results for the Irradiation Effects Test Series IE-ST-2, Rod IE-002

International Nuclear Information System (INIS)

Murdock, B.A.

1977-12-01

A postirradiation examination was conducted on a zircaloy-clad, UO 2 -fueled, pressurized water reactor (PWR) type rod which had been tested in the Power Burst Facility as part of the Irradiation Effects Test Series of the Thermal Fuels Behavior Program. The fuel rod, previously irradiated to a burnup of 15,800 MWd/t was subjected to a power ramp from 28 to 55 kW/m peak power at an average ramp rate of 4 kW/m/min. Posttest fuel restructuring and relocation, fission product redistribution, and fuel rod cladding deformation were evaluated and analyzed

Estimation of the core-wide fuel rod damage during a LWR LOCA

International Nuclear Information System (INIS)

Mattila, L.; Sairanen, R.; Stengaard, J.-O.

1975-01-01

The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

International Nuclear Information System (INIS)

Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

1982-01-01

The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

Investigation of control rod worth and nuclear end of life of BWR control rods

International Nuclear Information System (INIS)

Magnusson, Per

2008-01-01

This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10 B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10 B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10 B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

High-energy azimuthally polarized laser beam generation from an actively Q-switched Nd:YAG laser with c-cut YVO4 crystal

Science.gov (United States)

Guo, Jing; Zhang, Baofu; Jiao, Zhongxing; He, Guangyuan; Wang, Biao

2018-05-01

A high-energy, azimuthally polarized (AP) and actively Q-switched Nd:YAG laser is demonstrated. The thermal bipolar lensing effect in the Nd:YAG laser rod is used as a polarization discriminator, and a c-cut YVO4 crystal is inserted into the laser cavity to increase the mode-selecting ability of the cavity for AP mode. The laser generated AP pulses with maximum pulse energy as high as 4.2 mJ. To the best of our knowledge, this is the highest pulse energy obtained from an actively Q-switched AP laser. The pulse energy remained higher than 1 mJ over a wide range of repetition rates from 5 kHz to 25 kHz.

Single-phase convective heat transfer in rod bundles

International Nuclear Information System (INIS)

Holloway, Mary V.; Beasley, Donald E.; Conner, Michael E.

2008-01-01

The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids

Single-phase convective heat transfer in rod bundles

Energy Technology Data Exchange (ETDEWEB)

Holloway, Mary V. [Mechanical Engineering Department, United States Naval Academy, 590 Holloway Rd., Annapolis, MD 21402 (United States)], E-mail: holloway@usna.edu; Beasley, Donald E. [Mechanical Engineering Department, Clemson University, Clemson, SC 29634 (United States); Conner, Michael E. [Westinghouse Nuclear Fuel, 5801 Bluff Road, Columbia, SC 29250 (United States)

2008-04-15

The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids.

Fabrication of internally instrumented reactor fuel rods

International Nuclear Information System (INIS)

Schmutz, J.D.; Meservey, R.H.

1975-01-01

Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

Vortex Noise from Rotating Cylindrical Rods

Science.gov (United States)

Stowell, E Z; Deming, A F

1935-01-01

A series of round rods of the some diameter were rotated individually about the mid-point of each rod. Vortices are shed from the rods when in motion, giving rise to the emission of sound. With the rotating system placed in the open air, the distribution of sound in space, the acoustical power output, and the spectral distribution have been studied. The frequency of emission of vortices from any point on the rod is given by the formula von Karman. From the spectrum estimates are made of the distribution of acoustical power along the rod, the amount of air concerned in sound production, the "equivalent size" of the vortices, and the acoustical energy content for each vortex.

F.E.M. of PWR's control rod cluster. Parametrical study of vibrating behavior by an Experiment Design method

International Nuclear Information System (INIS)

Bosselut, D.; Soulier, B.

1997-03-01

Some finite element models have been performed at EDF to simulate the vibrations of rod cluster and to analyse the wear phenomenon of rods using parametrical studies. In the first part, one of the finite element models is presented. The location of excitation sources is described. The calculated values are: rod displacement in the guiding cards, shock forces on the guiding cards and wear power produced. In the second part, a parametrical study is presented for a given computer experiment domain with an Experimental Design method. The building of the computer experiment design is described. The used polynomial model has all linear, quadratic and interactive terms for each of the 6 parameters (26 coefficients), 34 polynomials have been built to approach the effective shock forces and the mean wear power at each of the 17 guiding points. In the last part, the influence of parameters on calculated mean wear power is shown along rods and some responses surfaces are visualized. Systematism and closeness of experiment design technique is underlined. Easy simulation of all the response domain by polynomial approach, allows comparison with experiment feedback. (author)

Flow field measurements using LDA and numerical computation for rod bundle of reactor fuel assembly

International Nuclear Information System (INIS)

Hu Jun; Zou Zunyu

1995-02-01

Local mean velocity and turbulence intensity measurements were obtained with DANTEC 55 X two-dimensional Laser Dopper Anemometry (LDA) for rod bundle of reactor fuel assembly test model which was a 4 x 4 rod bundle. The data were obtained from different experimental cross-sections both upstream and downstream of the model support plate. Measurements performed at test Reynolds numbers of 1.8 x 10 4 ∼3.6 x 10 4 . The results described the local and gross effects of the support plate on upstream and downstream flow distributions. A numerical computation was also given, the experimental results are in good agreement with the numerical one and the others in references. Finally, a few suggestions were proposed for how to use the LDA system well. (11 figs.)

Air velocity profiles near sleeve blockages in an unheated 7 x 7 rod bundle. [PWR

Energy Technology Data Exchange (ETDEWEB)

Creer, J. M.; Bates, J. M.

1979-04-01

Local air velocity measurements were obtained with a laser Doppler anemometer near flow blockages in an unheated 7 x 7 rod bundle. Sleeve blockages were positioned on the center nine rods to create an area reduction of 90% in the center four subchannels of the bundle. Experimental results indicated that severe flow disturbances occurred downstream from the blockage cluster but showed only minor flow disturbances upstream from the blockage. Flow reversals were detected downstream from the blockage and persisted for approximately five subchannel hydraulic diameters. The air velocity profiles were in excellent agreement with water velocity data previously obtained at essentially the same Reynolds number. Subchannel average velocity predictions obtained with the COBRA computer program were in good agreement with subchannel average velocities estimated using the measured local velocity data.

Fuel and control rod failure behavior during degraded core accidents

International Nuclear Information System (INIS)

Chung, K.S.

1984-01-01

As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

Fabrication of control rods for the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Sease, J.D.

1998-01-01

The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A

Fabrication of control rods for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Sease, J.D.

1998-03-01

The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

Design of ultrahigh brightness solar-pumped disk laser.

Science.gov (United States)

Liang, Dawei; Almeida, Joana

2012-09-10

To significantly improve the solar-pumped laser beam brightness, a multi-Fresnel lens scheme is proposed for side-pumping either a single-crystal Nd:YAG or a core-doped ceramic Sm(3+) Nd:YAG disk. Optimum laser system parameters are found through ZEMAX and LASCAD numerical analysis. An ultrahigh laser beam figure of merit B of 53 W is numerically calculated, corresponding to a significant enhancement of more than 180 times over the previous record. 17.7 W/m(2) collection efficiency is also numerically attained. The strong thermal effects that have hampered present-day rod-type solar-pumped lasers can also be largely alleviated.

Vibrational characteristics and wear of fuel rods

International Nuclear Information System (INIS)

Schmugar, K.L.

1977-01-01

Fuel rod wear, due to vibration, is a continuing concern in the design of liquid-cooled reactors. In my report, the methodology and models that are used to predict fuel rod vibrational response and vibratory wear, in a light water reactor environment, are discussed. This methodology is being followed at present in the design of Westinghouse Nuclear Fuel. Fuel rod vibrations are expressed as the normal bending modes, and sources of rod vibration are examined with special emphasis on flow-induced mechanisms in the stable flow region. In a typical Westinghouse PWR fuel assembly design, each fuel rod is supported at multiple locations along the rod axis by a square-shaped 'grid cell'. For a fuel rod /grid support system, the development of small oscillatory motions, due to fluid flow at the rod/grid interface, results in material wear. A theoretical wear mode is developed using the Archard Theory of Adhesive Wear as the basis. Without question certainty, fretting wear becomes a serious problem if it progresses to the stage where the fuel cladding is penetrated and fuel is exposed to the coolant. Westinghouse fuel is designed to minimize fretting wear by limiting the relative motion between the fuel rod and its supports. The wear producing motion between the fuel rod and its supports occurs when the vibration amplitude exceeds the slippage threshold amplitude

Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

International Nuclear Information System (INIS)

2010-01-01

The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

Airclad fiber laser technology

DEFF Research Database (Denmark)

Hansen, Kim P.; Olausson, Christina Bjarnal Thulin; Broeng, Jes

2008-01-01

High-power fiber lasers and amplifiers have gained tremendous momentum in the last five years, and many of the traditional manufactures of gas and solid-state lasers are pursuing the attractive fiber-based systems, which are now displacing the old technology in many areas. High-power fiber laser...... systems require specially designed fibers with large cores and good power handling capabilities - requirements that are all met by the airclad fiber technology. In the present paper we go through many of the building blocks needed to build high-power systems and we show an example of a complete airclad...... laser system. We present the latest advancements within airclad fiber technology including a new 70 μm single-mode polarization-maintaining rod-type fiber capable of amplifying to MW power levels. Furthermore we describe the novel airclad based pump combiners and their use in a completely monolithic 350...

Radial power density distribution of MOX fuel rods in the IFA-651

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung Ho; Koo, Yang Hyun; Joo, Hyung Kook; Cheon, Jin Sik; Oh, Je Yong; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

Two MOX fuel rods, which were fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with Korea Atomic Energy Research Institute, have been irradiated in the HBWR from June, 2000 in the framework of OECD-HRP together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is basic in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR{sub H}BWR that calculates radial power density distribution for three MOX fuel rods has been developed based on neutron physics results and DEPRESS program. The developed subroutine FACTOR{sub H}BWR gives good agreement with the physics calculation except slight under-prediction at the outer part of the pellet above the burnup of 20 MWd/kgHM. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. 24 figs., 4 tabs. (Author)

Drive-in device for long thin rods into narrow cavitations, especially for control-shutdown rods e.g. of nuclear reactors

International Nuclear Information System (INIS)

Flessner, H.; Paeserack, U.

1974-01-01

The auxiliary device serves as holder for long and thin rods, e.g. control rods, transported hanging in bundles, when these are lowered into narrow cavities. It is constructed as a rod grab vertically movable at the end of a guide tube. A comb-shaped trap in connection with a guide rod serves for lateral support of the lower ends of the rods hanging on the grab. This guide rod can be moved in vertical direction by means of two pairs of convex rollers resting on the inner guide tube. In addition, the guide rod has a prolongation carrying a traverse by means of an abutment on the lower end. With these auxiliaries amongst others, the trap can be brought into a horizontal position by turning around an axis with the control rods meshing with the teeth of the trap while the parallelism of the rods is kept up during transport. (DG) [de

Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

International Nuclear Information System (INIS)

Siefken, L.J.

1999-01-01

Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from above on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown

Control rod excess withdrawal prevention device

International Nuclear Information System (INIS)

Takayama, Yoshihito.

1992-01-01

Excess withdrawal of a control rod of a BWR type reactor is prevented. That is, the device comprises (1) a speed detector for detecting the driving speed of a control rod, (2) a judging circuit for outputting an abnormal signal if the driving speed is greater than a predetermined level and (3) a direction control valve compulsory closing circuit for controlling the driving direction of inserting and withdrawing a control rod based on an abnormal signal. With such a constitution, when the with drawing speed of a control rod is greater than a predetermined level, it is detected by the speed detector and the judging circuit. Then, all of the direction control valve are closed by way of the direction control valve compulsory closing circuit. As a result, the operation of the control rod is stopped compulsorily and the withdrawing speed of the control rod can be lowered to a speed corresponding to that upon gravitational withdrawal. Accordingly, excess withdrawal can be prevented. (I.S)

Hollow rods for the oil producing industry

Energy Technology Data Exchange (ETDEWEB)

Khalimova, L M; Elyasheva, M A

1970-01-01

Hollow sucker rods have several advantages over conventional ones. The hollow rods actuate the well pump and at the same time conduct produced fluids to surface. When paraffin deposition occurs, it can be minimized by injecting steam, hot oil or hot water into the hollow rod. Other chemicals, such as demulsifiers, scale inhibitors, corrosion inhibitors, etc., can also be placed in the well through the hollow rods. This reduces cost of preventive treatments, reduces number of workovers, increases oil production, and reduces cost of oil. Because the internal area of the rod is small, the passing liquids have a high velocity and thereby carry sand and dirt out of the well. This reduces pump wear between the piston and the plunger. Specifications of hollow rods, their operating characteristics, and results obtained with such rods under various circumstances are described.

Mechanical behaviour of PWR fuel rods during intermediate storage

International Nuclear Information System (INIS)

Bouffioux, P.; Dalmas, R.; Bernaudat, C.

2000-01-01

EDF, which owns the irradiated fuel coming from its NPPs, has initiated studies regarding the mechanical behaviour of a fuel rod and the integrity of its cladding, in the case where the spent fuel is stored for a significant duration. During the phases following in-reactor irradiation (ageing in a water-pool, transport and intermediate storage), many phenomena, which are strongly coupled, may influence the cladding integrity: - residual power and temperature decay; - helium production and release in the free volume of the rod (especially for MOX fuel); - fuel column swelling; - cladding creep-out under the inner gas pressure of the fuel rod; - metallurgical changes due to high temperatures during transportation. In parallel, the quantification of the radiological risk is based on the definition of a cladding integrity criterion. Up to now, this criterion requires that the clad hoop strain due to creep-out does not exceed 1%. A more accurate criterion is being investigated. The study and modelling of all the phenomena mentioned above are included in a R and D programme. This programme also aims at redefining the cladding integrity criterion, which is assumed to be too conservative. The R and D programme will be presented. In order to predict the overall behaviour of the rod during the intermediate storage phases, the AVACYC code has been developed. It includes the models developed in the R and D programme. The input data of the AVACYC code are provided by the results of in-reactor rod behaviour simulations, using the thermal-mechanical CYRANO3 code. Its main results are the evolution vs. time of hoop stresses in the cladding, rod internal pressure and cladding hoop strains. Chained CYRANO-AVACYC calculations have been used to simulate the behaviour of MOX fuel rods irradiated up to 40 GWd/t and stored under air during 100 years, or under water during 50 years. For such fuels, where the residual power remains high, we show that a large part of the cladding strain

Comparison of control rod effectiveness for thorium and low-enriched fuel cycles in the GA-1, 160 MW(e) design

Energy Technology Data Exchange (ETDEWEB)

Neef, Hans Joachim

1974-03-15

In an investigation of the properties of the Thorium-Uranium (Th) and the Low-Enriched Uranium (LEU) fuel cycles it is also necessary to compare the effectiveness of the control rods in a reactor system operating with these sorts of fuel. Furthermore, it is under consideration to start a reactor with LEU fuel and switch-over to a Th cycle. It is also of interest to look at the switch-over phase in respect to the control rod effectiveness. The various fuel cycles have been studied for the same fuel element and control rod design, namely the one of GA's commercially available 1,160 MW(e) reference power station. This paper gives the first results on the control rod calculations and is presented mainly in two parts. Part 1 describes spectral effects which have been investigated by cell calculations with a discrete ordinates transport code. The main result is the higher effectiveness of a rod in a Th-cycle compared with a LEU-cycle. Part 2 reports on reactor calculations with a diffusion code and shows that this advantage can partially disappear in the reactor because of the spatial flux distribution. This effect has to be studied in further investigations for a full understanding.

Rolls-Royce digital Rod Control System

International Nuclear Information System (INIS)

Pouillot, M.

2010-01-01

Full text of publication follows: Rolls-Royce has developed a new generation of Rod Control System, based on 40 years of experience. The fifth-generation Rod Control System (RCS) from Rolls-Royce offers a reliable, modular design with adaptability to your preferred platform, for modernization projects or new reactors. Flexible implementation provides the option for you to keep existing cabinets, which permits you to optimize installation approach. Main features for the power part: - Control Rod Drive Mechanism (CRDM) type: 3-coil. - Independent control of each sub-bank. - Each sub-bank is controlled by a cycler unit and 3 identical power racks, each including 4 identical power modules and a common power-supply module. - Coil-per-coil digital control: each power module embeds power-conversion, current-control, and current-monitoring functions for one coil. Control and monitoring are carried out by separate electronics in the module. Current is digitized and fully monitored by means of min-max templates. - A double-hold function is included: a power module assigned to a gripper will activate its coil if a fault risking to cause a reactor trip occurs. - Power modules are standardized, hot-pluggable and self-configured: a power module includes a set of parameters for each type of coil SG, MG, LC. The module recognizes the rack it is plugged in, and chooses automatically parameters to be used. Main benefits: - Reduced operational, maintenance, training, and inventory costs: standardization of power modules and integration of control and monitoring on the same PC-card lead to a drastic reduction of spare part types, and simplification of the system. - Easy maintenance: - Replacement of a power module solves nearly all failures due to current control or monitoring for a coil. It is done instantly thanks to hot-plug capability. - On the front plate of power-modules, LEDs provide useful information for diagnostic: current setpoint from cycler, output current bar

ELECTRIC FIELD MEASUREMENT IN ROD-DISCONTINUED ...

African Journals Online (AJOL)

2014-06-30

Jun 30, 2014 ... the electrogeometrical model using a laboratory experimental rod-plane air gap arrangement with a lightning conductor (Franklin rod or horizontal conductor). The stepped leader could be represented by the rod electrode under a negative lightning impulse voltage having a level leading to breakdown with ...

Method of inspecting control rod drive mechanism

International Nuclear Information System (INIS)

Sato, Tomomi; Tatemichi, Shin-ichiro; Hasegawa, Hidenobu.

1988-01-01

Purpose: To conduct inspection for control rod drives and fuel handling operations in parallel without taking out the entire fuel, while maintaining the reactor in a subcritical state. Method: Control rod drives are inspected through the release of connection between control rods and control rod drives, detachment and dismantling of control rod drives, etc. In this case, structural materials having neutron absorbing power equal to or greater than the control rods are inserted into the gap after taking out fuels. Since the structural materials have neutron absorbing portion, subcriticality is maintained by the neutron absorbing effect. Accordingly, there is no requirement for taking out all of the fuels, thereby enabling to check the control rod drives and conduct handling for the fuels in parallel. As a result, the number of days required for the inspection can be shortened and it is possible to improve the working efficiency for the decomposition, inspection, etc. of the control rod drives and, thus, improve the operation efficiency of the nuclear power plant thereby attaining the predetermined purpose. (Kawakami, Y.)

Control rod

International Nuclear Information System (INIS)

Takeda, Toshikazu; Inoue, Kotaro.

1979-01-01

Purpose: To flatten the power distribution in the reactor core without impairing neutron economy by disposing pins containing elements of lower atomic number in the central region of a shroud and loading pins containing depleted uranium in the periphery region thereof. Constitution: The shroud has a layer of pins containing depleted uranium in the peripheral region and a layer of pins containing elements of lower atomic number such as beryllium in the central region. Heat removal from those pins containing depleted uranium and elements of lower atomic number (neutron moderator) is effected by sodium flow outside of the cladding material. The control rod operation is conducted by inserting or extracting the central portion (pins containing elements of lower atomic number such as beryllium) inside of the stainless pipe. Upon extraction of the control rod, the moderator in the central region is removed whereby high speed neutrons are no more deccelerated and the absorption rate to the depleted uranium is decreased. This can flatten the power distribution in the reactore core with the disposition of a plurality of control rods at a better neutron economy as compared with the use of neutron absorber such as boron. (Seki, T.)

Control rod

International Nuclear Information System (INIS)

Fukumoto, Takashi; Hirakawa, Hiromasa; Kawashima, Norio; Goto, Yasuyuki.

1994-01-01

Neutron absorbers are contained in a tubular member comprising, integrally a tubular portion and four corners disposed at the outer circumference of the tubular portion at every 90deg, to provide a neutron absorbing tube. A plurality of neutron absorbing tubes are arranged in parallel in the lateral direction, and adjacent corners are joined, into a blade to constitute a control rod. Such a control rod has a great structural strength, simple in the structure and relatively light in weight and can contain a great amount of neutron absorbers. Upon formation of the control rod by arranging the blades in a cross-like shape, at least a portion thereof is constituted with short neutron absorbing tubes shorter than the entire length of the blade, and gaps are formed at positions in adjacent in the axial direction. With such a constitution, there is no worry that a wing end of the blade collides against or be abraded with a fuel channel box or a fuel support. Even if fuel channels are vibrated upon scram of the reactor, such as occurrence of earthquakes, it can be inserted to the reactor easily. (N.H.)

Experimental determination of temperature fields in sodium-cooled rod bundles with hexagonal rod arrangement and grid spacers

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.; Kolodziej, M.

1977-01-01

Three-dimensional temperature fields in the claddings of sodium cooled rods were determined experimentally under representative nominal operating conditions for a SNR typical 19-rod bundle model provided with spark-eroded spacers. These experiments are required to verify thermohydraulic computer programs which will provide the output data for strength calculations of the high loaded cladding tubes. In this work the essentials are reported of the measured circumferential distributions of wall temperatures of peripheral rods. In addition the sub-channel temperatures measured over the bundle cross section are indicated, they are required to sustain codes for the global thermohydraulic design of core elements. The most important results are: 1) The whole fuel element is located within the thermal entrance length. 2) High azimuthal temperature differences were measured in the claddings of peripheral rods, which are strongly influenced by the distance between the rod and the shroud, especially for the corner rod. 3) With decreasing Pe-number ( [de

Control rod housing alignment

International Nuclear Information System (INIS)

Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

1990-01-01

This patent describes a process for measuring the vertical alignment between a hole in a core plate and the top of a corresponding control rod drive housing within a boiling water reactor. It comprises: providing an alignment apparatus. The alignment apparatus including a lower end for fitting to the top of the control rod drive housing; an upper end for fitting to the aperture in the core plate, and a leveling means attached to the alignment apparatus to read out the difference in angularity with respect to gravity, and alignment pin registering means for registering to the alignment pin on the core plate; lowering the alignment device on a depending support through a lattice position in the top guide through the hole in the core plate down into registered contact with the top of the control rod drive housing; registering the upper end to the sides of the hole in the core plate; registering the alignment pin registering means to an alignment pin on the core plate to impart to the alignment device the required angularity; and reading out the angle of the control rod drive housing with respect to the hole in the core plate through the leveling devices whereby the angularity of the top of the control rod drive housing with respect to the hole in the core plate can be determined

The Third ATLAS ROD Workshop

CERN Multimedia

Poggioli, L.

A new-style Workshop After two successful ATLAS ROD Workshops dedicated to the ROD hardware and held at the Geneva University in 1998 and in 2000, a new style Workshop took place at LAPP in Annecy on November 14-15, 2002. This time the Workshop was fully dedicated to the ROD-TDAQ integration and software in view of the near future integration activities of the final RODs for the detector assembly and commissioning. More precisely, the aim of this workshop was to get from the sub-detectors the parameters needed for T-DAQ, as well as status and plans from ROD builders. On the other hand, what was decided and assumed had to be stated (like EB decisions and URDs), and also support plans. The Workshop gathered about 70 participants from all ATLAS sub-detectors and the T-DAQ community. The quite dense agenda allowed nevertheless for many lively discussions, and for a dinner in the old town of Annecy. The Sessions The Workshop was organized in five main sessions: Assumptions and recommendations Sub-de...

Temperature actuated automatic safety rod release

Science.gov (United States)

Hutter, E.; Pardini, J.A.; Walker, D.E.

1984-03-13

A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

Shielding device for control rod in nuclear reactor

International Nuclear Information System (INIS)

Sakamaki, Kazuo; Tomatsu, Tsutomu.

1995-01-01

The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

Burnable poison rod

International Nuclear Information System (INIS)

Natsume, Tomohiro.

1988-01-01

Purpose: To increase the reactor core lifetime by decreasing the effect of neutron absorption of burnable poison rods by using material with less neutron absorbing effect. Constitution: Stainless steels used so far as the coating material for burnable poison rods have relatively great absorption in the thermal neutral region and are not preferred in view of the neutron economy. Burnable poison rods having fuel can made of zirconium alloy shows absorption the thermal neutron region lower by one digit than that of stainless steels but they shows absorption in the resonance region and the cost is higher. In view of the above, the fuel can of the burnable poison material is made of aluminum or aluminu alloy. This can reduce the neutron absorbing effect by stainless steel fuel can and effectively utilize neutrons that have been wastefully absorbed and consumed in stainless steels. (Takahashi, M.)

The GEO 600 laser system

CERN Document Server

Zawischa, I; Danzmann, K; Fallnich, C; Heurs, M; Nagano, S; Quetschke, V; Welling, H; Willke, B

2002-01-01

Interferometric gravitational wave detectors require high optical power, single frequency lasers with very good beam quality and high amplitude and frequency stability as well as high long-term reliability as input light source. For GEO 600 a laser system with these properties is realized by a stable planar, longitudinally pumped 12 W Nd:YAG rod laser which is injection-locked to a monolithic 800 mW Nd:YAG non-planar ring oscillator. Frequency control signals from the mode cleaners are fed to the actuators of the non-planar ring oscillator which determines the frequency stability of the system. The system power stabilization acts on the slave laser pump diodes which have the largest influence on the output power. In order to gain more output power, a combined Nd:YAG-Nd:YVO sub 4 system is scaled to more than 22 W.

Progress in solid state dye laser development

Energy Technology Data Exchange (ETDEWEB)

Hermes, R.E.

1990-01-01

A triaxial flashlamp (15 cm) was used to optically pump laser rods prepared from an acrylate based copolymer (0.95 cm O.D. {times} 10.0 cm L.). The performance of 13 laser dyes incorporated into this polymeric solid host is reported. The best lasing performance was obtained with sulforhodamine-B, with a calculated slope efficiency of 0.52% and a maximum single pulse output energy of 580 mJ. A commercially available fluorescent polymeric material was also evaluated. 12 refs., 2 figs.

Thermal expansion measurements on boron carbide and europium sesquioxide by laser interferometry

International Nuclear Information System (INIS)

Preston, S.D.

1980-01-01

A laser interferometer technique for measuring the absolute linear thermal expansion of small annular specimens is described. Results are presented for unirradiated boron carbide (B 4 C) and europia (Eu 2 O 3 ) up to 1000 0 C. Both compounds are neutron-absorbing materials of potential use in fast-reactor control rods and data on their thermophysical properties, in particular linear thermal expansion, are essential to the control rod designers. (author)

Simulation of leaking fuel rods

International Nuclear Information System (INIS)

Hozer, Z.

2006-01-01

The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123 I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123 I release from failed fuel rods during transients

Control rod for HTGR type reactor

International Nuclear Information System (INIS)

Mogi, Haruyoshi; Saito, Yuji; Fukamichi, Kenjiro.

1990-01-01

Upon dropping control rod elements into the reactor core, impact shocks are applied to wire ropes or spines to possibly deteriorate the integrity of the control rods. In view of the above in the present invention, shock absorbers such as springs or bellows are disposed between a wire rope and a spine in a HTGR type reactor control rod comprising a plurality of control rod elements connected axially by means of a spine that penetrates the central portion thereof, and is suspended at the upper end thereof by a wire rope. Impact shocks of about 5 kg are applied to the wire rope and the spine and, since they can be reduced by the shock absorbers, the control rod integrity can be maintained and the reactor safety can be improved. (T.M.)

Why Rods and Cocci

Indian Academy of Sciences (India)

Bacteria exhibit a wide variety of shapes but the commonly studied species of bacteria are generally either spherical in shape which are called cocci (singular coccus) or have a cylindrical shape and are called rods or bacilli (singular bacillus). In reality rods and cocci are the ends of a continuum. Sonle of the cocci are.

Spacers for fuel rod clusters

International Nuclear Information System (INIS)

Jabsen, F.S.

1978-01-01

The proposition deals with the fixing of nuclear fuel element rods in a grid which consists of a number of crossed Zy-plates which form cells. The rectangular cells have projections which serve as spacers for the fuel rods. According to the invention there are additional butt straps which can be moved in such a way that insertion and extraction of the fuel rods can be done without obstruction and they can be spring-loaded hold in their final position. (UWI) [de

Rod consolidation at the West Valley Demonstration Project

International Nuclear Information System (INIS)

Bailey, W.J.

1986-12-01

A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab

21 CFR 876.4270 - Colostomy rod.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out through...

Control rod for FBR type reactor

International Nuclear Information System (INIS)

Nakai, Koichi.

1993-01-01

In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

Design of active-neutron fuel rod scanner

International Nuclear Information System (INIS)

Griffith, G.W.; Menlove, H.O.

1996-01-01

An active-neutron fuel rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A 252 Cf source is located at the center of the scanner very near the through hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. We used the Monte Carlo code MCNP to design the scanner and review optimum materials and geometries. An inhomogeneous beryllium, graphite, and polyethylene moderator has been designed that uses source neutrons much more efficiently than assay systems using polyethylene moderators. Layers of borated polyethylene and tungsten are used to shield the detectors. Large NaI(Tl) detectors were selected to measure the delayed gamma rays. The enrichment zones of a thermal reactor fuel pin could be measured to within 1% counting statistics for practical rod speeds. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels. (orig.)

Hydraulically centered control rod

International Nuclear Information System (INIS)

Horlacher, W.R.; Sampson, W.T.; Schukei, G.E.

1981-01-01

A control rod suspended to reciprocate in a guide tube of a nuclear fuel assembly has a hydraulic bearing formed at its lower tip. The bearing includes a plurality of discrete pockets on its outer surface into which a flow of liquid is continuously provided. In one embodiment the flow is induced by the pressure head in a downward facing chamber at the end of the bearing. In another embodiment the flow originates outside the guide tube. In both embodiments the flow into the pockets produces pressure differences across the bearing which counteract forces tending to drive the rod against the guide tube wall. Thus contact of the rod against the guide tube is avoided

In-chip microstructures and photonic devices fabricated by nonlinear laser lithography deep inside silicon

Science.gov (United States)

Tokel, Onur; Turnalı, Ahmet; Makey, Ghaith; Elahi, Parviz; ćolakoǧlu, Tahir; Ergeçen, Emre; Yavuz, Ã.-zgün; Hübner, René; Zolfaghari Borra, Mona; Pavlov, Ihor; Bek, Alpan; Turan, Raşit; Kesim, Denizhan Koray; Tozburun, Serhat; Ilday, Serim; Ilday, F. Ã.-mer

2017-10-01

Silicon is an excellent material for microelectronics and integrated photonics1-3, with untapped potential for mid-infrared optics4. Despite broad recognition of the importance of the third dimension5,6, current lithography methods do not allow the fabrication of photonic devices and functional microelements directly inside silicon chips. Even relatively simple curved geometries cannot be realized with techniques like reactive ion etching. Embedded optical elements7, electronic devices and better electronic-photonic integration are lacking8. Here, we demonstrate laser-based fabrication of complex 3D structures deep inside silicon using 1-µm-sized dots and rod-like structures of adjustable length as basic building blocks. The laser-modified Si has an optical index different to that in unmodified parts, enabling the creation of numerous photonic devices. Optionally, these parts can be chemically etched to produce desired 3D shapes. We exemplify a plethora of subsurface—that is, `in-chip'—microstructures for microfluidic cooling of chips, vias, micro-electro-mechanical systems, photovoltaic applications and photonic devices that match or surpass corresponding state-of-the-art device performances.

In-chip microstructures and photonic devices fabricated by nonlinear laser lithography deep inside silicon.

Science.gov (United States)

Tokel, Onur; Turnali, Ahmet; Makey, Ghaith; Elahi, Parviz; Çolakoğlu, Tahir; Ergeçen, Emre; Yavuz, Özgün; Hübner, René; Borra, Mona Zolfaghari; Pavlov, Ihor; Bek, Alpan; Turan, Raşit; Kesim, Denizhan Koray; Tozburun, Serhat; Ilday, Serim; Ilday, F Ömer

2017-10-01

Silicon is an excellent material for microelectronics and integrated photonics1-3 with untapped potential for mid-IR optics4. Despite broad recognition of the importance of the third dimension5,6, current lithography methods do not allow fabrication of photonic devices and functional microelements directly inside silicon chips. Even relatively simple curved geometries cannot be realised with techniques like reactive ion etching. Embedded optical elements, like in glass7, electronic devices, and better electronic-photonic integration are lacking8. Here, we demonstrate laser-based fabrication of complex 3D structures deep inside silicon using 1 µm-sized dots and rod-like structures of adjustable length as basic building blocks. The laser-modified Si has a different optical index than unmodified parts, which enables numerous photonic devices. Optionally, these parts are chemically etched to produce desired 3D shapes. We exemplify a plethora of subsurface, i.e. , " in-chip" microstructures for microfluidic cooling of chips, vias, MEMS, photovoltaic applications and photonic devices that match or surpass the corresponding state-of-the-art device performances.

Heat flux and temperature determination on the control rod outer surface

International Nuclear Information System (INIS)

Taler, J.; Cebula, A.; Marcinkiewicz, J.; Tinoco, H.

2011-01-01

The paper presents heat transfer calculation results concerning a control rod of Unit 3 of Forsmark Nuclear Power Plant (NPP). The part of the control rod, which is the object of interest, is surrounded by a mixing region of hot and cold flows and, as a consequence, is subjected to thermal fluctuations. The paper describes a numerical test which validates the method based on the solution of the inverse heat conduction problem (IHCP). The comparison of the results achieved by two methods, CFD and IHCP, including a description of the IHCP method used in the calculation process, shows a very good agreement between the methods. (author)

Spider and burnable poison rod combinations

International Nuclear Information System (INIS)

Edwards, G.T.; Schluderberg, D.C.

1980-01-01

An improved design of burnable poison rods and associated spiders used in fuel assemblies of pressurized water power reactor cores, is described. The rods are joined to the spider arms in a manner which is proof against the reactor core environment and yet allows the removal of the rods from the spider simply, swiftly and delicately. (U.K.)

Micromachining of laser fusion target parts

International Nuclear Information System (INIS)

Weir, J.T.; Hendricks, C.D.; Weinstein, B.W.; Willenborg, D.L.

1976-01-01

A 5W argon ion laser that operates CW is used. A broad band rear mirror is tuned to maximum power output. The beam is directed vertically by an adjustable turning prism, through a beam splitter, and then focused with an ordinary microscope objective lens onto the material to be cut. The beam splitter allows a telescope and television camera arranged to view the cutting through the same lens that is focusing the laser. The object to be cut is mounted on a micromanipulator which can move the object in two dimensions in the focal plane of the laser

Laser surveillance system for spent fuel

International Nuclear Information System (INIS)

Fiarman, S.; Zucker, M.S.; Bieber, A.M. Jr.

1980-01-01

A laser surveillance system installed at spent fuel storage pools will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a laser system and have used it with a simulated BWR assembly. The reflected signal from the zircaloy rods depends on the position of the assembly, but in all cases is easily discernable from the reference scan of background with no assembly

Fuel element cellular grid structure and procedure to insert and withdraw fuel rods from that structure

International Nuclear Information System (INIS)

1975-01-01

A typical embodiment of the invention provides a means for selectively inserting and withdrawing one or more fuel rods from a fuel element cellular grid structure. The transverse stubs on one side of a long, thin bar are turned through 90deg to extend across the gap between mutually perpendicular grid structure plates. The extreme ends of these stubs engage the adhacent portions of the associated plates that form part of the grid cells. Pressing the stubs against the plate portions through the application of appropriate force in a longitudinal direction relative to the bar deflects the engaged plates through a sufficient distance to enable fuel rods to be inserted into or withdrawn from respective cells. After rod insertion, the force applied to the bar is released to enable the plates to relax and engage the fuel rods. The bars are rotated once more through 90deg and withdrawn from the grid structure. A similar procedure is employed to withdraw fuel rods from the grid structure

The development of laser ignited deflagration-to-detonation transition (DDT) detonators and pyrotechnic actuators

Energy Technology Data Exchange (ETDEWEB)

Merson, J.A.; Salas, F.J.; Harlan, J.G.

1993-11-01

The use of laser ignited explosive components has been recognized as a safety enhancement over existing electrical explosive devices (EEDs). Sandia has been pursuing the development of optical ordnance for many years with recent emphasis on developing optical deflagration-to-detonation (DDT) detonators and pyrotechnic actuators. These low energy optical ordnance devices can be ignited with either a semiconductor diode laser, laser diode arrays or a solid state rod laser. By using a semiconductor laser diode, the safety improvement can be made without sacrificing performance since the input energy required for the laser diode and the explosive output are similar to existing electrical systems. The use of higher powered laser diode arrays or rod lasers may have advantages in fast DDT applications or lossy optical environments such as long fiber applications and applications with numerous optical connectors. Recent results from our continued study of optical ignition of explosive and pyrotechnic materials are presented. These areas of investigation can be separated into three different margin categories: (1) the margin relative to intended inputs (i.e. powder performance as a function of laser input variation), (2) the margin relative to anticipated environments (i.e. powder performance as a function of thermal environment variation), and (3) the margin relative to unintended environments (i.e. responses to abnormal environments or safety).

Detection device for control rod interference

International Nuclear Information System (INIS)

Saito, Noboru.

1984-01-01

Purpose: To enable to detect the mechanical interference or friction between a control rod and a channel box automatically, simply and rapidly. Constitution: A signal from a gate circuit and a signal from a comparison mechanism are inputted into an AND circuit if a control rod has not been displaced by a predetermined distance within a prescribed time Δt after the output of an insertion or withdrawal signal for the control rod, by which a control-rod-interference signal is outputted from the AND circuit. Accordingly, the interference between the control rod and the channel box can be detected automatically, easily and rapidly. Furthermore, by properly adjusting the prescribed time Δt set by the gate circuit, the degree of the interference can also be detected, whereby the safety and the reliability of the reactor can be improved significantly. (Horiuchi, T.)

Sixteenth International Laser Radar Conference, Part 2

International Nuclear Information System (INIS)

Mccormick, M.P.

1992-07-01

Given here are extended abstracts of papers presented at the 16th International Laser Radar Conference, held in Cambridge, Massachusetts, July 20-24, 1992. Topics discussed include the Mt. Pinatubo volcanic dust laser observations, global change, ozone measurements, Earth mesospheric measurements, wind measurements, imaging, ranging, water vapor measurements, and laser devices and technology

Investigation of Different Colloidal Porous Silicon Solutions and Their Composite Solid Matrix Rods by Optical Techniques

Science.gov (United States)

Khan, M. Naziruddin; Aldalbahi, Ali; Almohammedi, Abdullah

2018-03-01

Colloidal porous silicon (PSi) in different solvents was synthesized by simple chemical etching. Colloidal solutions were then prepared using different quantities of silicon wafer pieces (Pcs) and chloroplatinic (Pt) acid in catalyst solution. The effect on the properties of the colloidal solutions and composite rods were investigated using various optical characterization techniques. Absorption and photoluminescence (PL) intensity of the colloidal PSi solutions are observed to depend on the quantity of wafer Pcs, the Pt-solution, and the porosity formation on the wafer surface. The morphological structure of the PSi in a solvent and the solid-rod environments were studied using field-emission scanning electron microscopy (FE-SEM) and were observed to have different structures. A mono-oriented structure of PSi exists in tetrahydrofuran, which has stereo orientation in dioxane and dimethylsulfoxide (approximately 5-8 nm as confirmed using high resolution transmission electron microscopy). Subsequently, some colloidal PSi solutions were directly embedded in three types of sol-gel-based matrices, silica, ormosils (or organically modified silica) and polymer, which easily generated solid rods. Spontaneous emission (SE) of the PSi solutions and their composite rods were examined using a high power picosecond 355 nm laser source. The emitted PL and SE signals of the colloidal PSi solutions were dependent on the Pt volume, nature of the solvent, quantity of Si wafer piece, and pumping energy. The response of SE signals from the PSi composites rods is an interesting phenomenon, and such nanocomposites may be used for future research on light amplification.

Laser surveillance system for spent fuel

International Nuclear Information System (INIS)

Fiarman, S.; Zucker, M.S.; Bieber, A.M. Jr.

1980-01-01

A laser surveillance system installed at spent fuel storage pools (SFSP's) will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a full size laser system operating in air and have used an array of 6 zircaloy BWR tubes to simulate an assembly. The reflective signal from the zircaloy rods is a strong function of position of the assembly, but in all cases is easily discernable from the reference scan of the background with no assembly. A design for a SFSP laser surveillance system incorporating laser ranging is discussed. 10 figures

Post test investigation of the single rod tests ESSI 1-11 on temperature escalation in PWR fuel rod simulators due to the Zircaloy/steam reaction

International Nuclear Information System (INIS)

Hagen, S.; Kapulla, H.; Malauschek, H.; Katanishi, S.

1987-03-01

This KfK-report describes the posttest investigation of the single rod tests ESSI-1 to ESSI-11. The objective of these tests was to investigate the temperature escalation behaviour of Zircaloy clad PWR-fuel rods in steam. The investigation of the temperature escalation is part of the program of out-of-pile experiments (CORA) performed within the frame work of the PNS Severe Fuel Damage Program. The experimental arrangement consisted of fuel rod simulator (central tungsten heater, UO 2 ring pellets and Zircaloy cladding), Zircaloy shroud and fiber ceramic insulation. The introductory test ESSI-1 to ESSI-3 were scoping tests designed to obtain information on the temperature escalation of zircaloy in steam. ESSI-4 to ESSI-8 were run with increasing heating rates to investigate the influence of the oxide layer thickness at the start of the escalation. ESSI-9 to ESSI-11 were performed to investigate the influence of the insulation thickness on the escalation behaviour. In these tests we also learned that the gap between removed shroud and insulation has a remarkable influence due to heat removal by convection in the gap. After the test the fuel rod simulator was embedded into epoxy and cut by a diamond saw. The cross sections were photographed and investigated by metalograph microscope, SEM and EMP examinations. (orig./GL) [de

Snubber assembly for a control rod drive

International Nuclear Information System (INIS)

1976-01-01

A snubber cartridge assembly is described which is mounted to the nozzle of a control rod drive mechanism to insure that it will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston-mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllable exhaust the liquid during a 'scram' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe 'scram' of the control rod into the reactor

Automated nuclear fuel rod pattern loading system

International Nuclear Information System (INIS)

Lambert, D.V.; Nyland, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

1990-01-01

This patent describes an apparatus for loading fuel rods in a desired pattern. It comprises: a carousel having a plurality of movable gondolas for stocking thereon fuel rods of known enrichments; an elongated magazine defining a matrix of elongated slots being open at their forward ends for receiving fuel rods; a workstation defining a fuel rod feed path; and a holder and indexing mechanism for movably supporting the magazine and being actuatable for moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

Snubber assembly for a control rod drive

International Nuclear Information System (INIS)

Matthews, J.C.

1978-01-01

A snubber cartridge assembly is mounted to the nozzle of a control rod drive mechanism to insure that the snubber assembly will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllably exhaust liquid therefrom during a ''scram'' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe ''scram'' of the control rod into the reactor

Method and apparatus for inspection of nuclear fuel rods

International Nuclear Information System (INIS)

Wachter, W.J.

1977-01-01

A method and apparatus are provided for the inspection of nuclear fuel rods to detect defects or failures in such rods. Assemblies of fuel rods are immersed in water and means are provided for causing a change in the relative pressures in the water and within the fuel rod such that fluid is expelled from the rod through any defects that may exist. Means are also provided for thereafter vibrating the rods to cause additional internal fluid or other material that may be trapped in the rod to be expelled. Sensors are provided for detecting the emission of bubbles of fluid or other material from the rod and for locating the position of the defective rod in the assembly. 5 figures

Mechanical reliability of falling systems. Case of control rods of nuclear reactors

International Nuclear Information System (INIS)

Rezkallah, B.

1991-05-01

After having briefly recalled the main principles adopted for nuclear reactor safety (barrier based method, emergency stop system, emergency cooling system) and how control rod operation anomalies are described, this research thesis first gives an overview of knowledge in the field of reliability (principles, qualitative and quantitative methods, failure trees, performance function, shock modelling and mode analysis). Then, the author presents the modelling of the drop and rotation of a control rod. The computation technique is the modal synthesis. The author highlights the influence of physical (friction, initial conditions) and numerical (modal truncation) parameters. He reports a probabilistic analysis of the problem and proposes a stochastic model for the different shock forces (notably friction). He compares the results with those obtained with the DEMT software. The last part reports the investigation of a rare case: the jamming or untimely stop of the control rod device

Mechanical properties of bioresorbable self-reinforced posterior cervical rods.

Science.gov (United States)

Savage, Katherine; Sardar, Zeeshan M; Pohjonen, Timo; Sidhu, Gursukhman S; Eachus, Benjamin D; Vaccaro, Alexander

2014-04-01

A biomechanical study. To test the mechanical and physical properties of self-reinforced copolymer bioresorbable posterior cervical rods and compare their mechanical properties to commonly used Irene titanium alloy rods. Bioresorbable instrumentation is becoming increasingly common in surgical spine procedures. Compared with metallic implants, bioresorbable implants are gradually reabsorbed as the bone heals, transferring the load from the instrumentation to bone, eliminating the need for hardware removal. In addition, bioresorbable implants produce less stress shielding due to a more physiological modulus of elasticity. Three types of rods were used: (1) 5.5 mm copolymer rods and (2) 3.5 mm and (3) 5.5 mm titanium alloy rods. Four tests were used on each rod: (1) 3-point bending test, (2) 4-point bending test, (3) shear test, and (4) differential scanning calorimeter test. The outcomes were recorded: Young modulus (E), stiffness, maximum load, deflection at maximum load, load at 1.0% strain of the rod's outer surface, and maximum bending stress. The Young modulus (E) for the copolymer rods (mean range, 6.4-6.8 GPa) was significantly lower than the 3.5 mm titanium rods (106 GPa) and the 5.5 mm titanium rods (95 GPa). The stiffness of the copolymer rods (mean range, 16.6-21.4 N/mm) was also significantly lower than the 3.5 mm titanium alloy rods (43.6 N/mm) and the 5.5 mm titanium alloy rods (239.6 N/mm). The mean maximum shear load of the copolymer rods was 2735 N and they had significantly lower mean maximum loads than the titanium rods. Copolymer rods have adequate shear resistance, but less load resistance and stiffness compared with titanium rods. Their stiffness is closer to that of bone, causing less stress shielding and better gradual dynamic loading. Their use in semirigid posterior stabilization of the cervical spine may be considered.

Implementation of CTRLPOS, a VENTURE module for control rod position criticality searches, control rod worth curve calculations, and general criticality searches

Energy Technology Data Exchange (ETDEWEB)

Smith, L.A.; Renier, J.P.

1994-06-01

A module in the VENTURE reactor analysis code system, CTRLPOS, is developed to position control rods and perform control rod position criticality searches. The module is variably dimensioned so that calculations can be performed with any number of control rod banks each having any number of control rods. CTRLPOS can also calculate control rod worth curves for a single control rod or a bank of control rods. Control rod depletion can be calculated to provide radiation source terms. These radiation source terms can be used to predict radiation doses to personnel and estimate the shielding and long-term storage requirements for spent control rods. All of these operations are completely automated. The numerous features of the module are discussed in detail. The necessary input data for the CTRLPOS module is explained. Several sample problems are presented to show the flexibility of the module. The results presented with the sample problems show that the CTRLPOS module is a powerful tool which allows a wide variety of calculations to be easily performed.

Absorber rod drive for nuclear reactors

International Nuclear Information System (INIS)

Acher, H.

1985-01-01

The invention concerns a further addition to the invention of DE 33 42 830 A1. The free contact of the hollow piston with the nut due to hydraulic pressure is replaced by a hydraulic or spring attachment. The pressure system required to produce the hydraulic pressure is therefore omitted, and the electrical power required for driving the pump or the mass flow is also omitted. The absorber rod slotted along its longitudinal axis is replaced by an absorber rod, in the longitudinal axis of which a hollow piston is connected together with the absorber rod. This makes the absorber rod more stable, and assembly is simplified. (orig./HP) [de

Nuclear reactor fuel rod attachment system

International Nuclear Information System (INIS)

Christiansen, D.W.

1983-01-01

The invention involves a technique to quickly, inexpensively and rigidly attach a nuclear reactor fuel rod to a support member. The invention also allows for the repeated non-destructive removal and replacement of the fuel rod. The proposed fuel rod and support member attachment and removal system consists of a locking cap fastened to the fuel rod and a locking strip fastened to the support member or vice versa. The locking cap has two or more opposing fingers shaped to form a socket. The fingers spring back when moved apart and released. The locking strip has an extension shaped to rigidly attach to the socket's body portion

New studies of the natural convection around a fuel rod of the BME training reactor with PIV/LIF technique

International Nuclear Information System (INIS)

Szijarto, R.; Aszodi, A.; Yamaji, B.

2011-01-01

In this paper the model of a fuel pin of the Training Reactor of Budapest University of Technology and Economics was investigated with Particle Image Velocimetry and Laser Induced Fluorescence measurement methods. An experimental setup was designed, built and optimized to investigate the natural convection around a model of a fuel pin of the Training Reactor. The processes were analysed using an electrically heated rod, which models the geometry of the fuel rods in the Training Reactor. The heated length of the model is the same as the active length of the real fuel rods. The rod is placed in a glass tank with a shape of a square-based prism. An additional cooling system ensures constant flow conditions around the rod. The setup consists of an additional flow channel box, the equivalent diameter of which is equal to the equivalent diameter of the real fuel assembly. Simultaneous measurements of velocity and temperature fields were performed in different vertical positions for both cases of natural convection with and without the flow channel box. The effect of the presence of the channel was analyzed, and a laminarizating influence was observed. The local heat transfer coefficient was calculated for every measurement. The two dimensional measurement techniques gave extensive results, the structure of the hydraulic and thermal boundary layer were fully analyzed. (Authors)

Comprehensive study of electro-optic and passive Q-switching in solid state lasers for altimeter applications

Science.gov (United States)

Bhardwaj, Atul; Agrawal, Lalita; Pal, Suranjan; Kumar, Anil

2006-12-01

Laser Science and Technology Center (LASTEC), Delhi, is developing a space qualified diode pumped Nd: YAG laser transmitter capable of generating 10 ns pulses of 30 mJ energy @ 10 pps. This paper presents the results of experiments for comparative studies between electro-optic and passively Q-switched Nd: YAG laser in a crossed porro prism based laser resonator. Experimental studies have been performed by developing an economical bench model of flash lamp pumped Nd: YAG laser (rod dimension, \

Control device for the withdrawal of control rod

International Nuclear Information System (INIS)

Ando, Masaki.

1985-01-01

Purpose: To significantly suppress the maximum value of the control-rod worth upon control rod withdrawal. Constitution: At first, a signal for designating the first class is sent from a class-control section to the group-control section. In the group-control section, the peripheral group among the first class is designated by which the withdrawal of the control rods other than the peripheral group is inhibited and the control-rods in the peripheral group are withdrawn one by one. When all of them have been withdrawn, the group-control section designates the central group of the first class. All the control rods of the central group have been withdrawn, then the group-control section designates the peripheral group of the second class. Thereafter, the central group in the second class is designated. The control rods are thus withdrawn in the same manner hereinafter. The maximum value for the control-rod worth can be decreased by such a withdrawing sequence for the control rods. (Horiuchi, T.)

Recent improvements in modelling fission gas release and rod deformation on metallic fuel in LMR

International Nuclear Information System (INIS)

Hwang, Woan; Lee, Byoung-Oon; Kim, Young Jin

2000-01-01

Metallic fuel design is a key feature to assure LMR core safety goals. To date, a large effort has been devoted to the development of the MACSIS code for metallic fuel rod design and the evaluation of operational limits under irradiation conditions. The updated models of fission gas release, fuel core swelling, and rod deformation are incorporated into the correspondence routines in MACSIS MOD1. The MACSIS MOD1 which is a new version of MACSIS, has been partly benchmarked on FGR, fuel swelling and rod deformation comparing with the results of U-Zr and U-Pu-Zr metal fuels irradiated in LMRs. The MACSIS MOD1 predicts, relatively well, the absolute magnitudes and trends of the gas release and rod deformations depending on burn-up, and it gives better agreement with the experimental data than the previous predictions of MACSIS and the results of the empirical model

Advances in Single-Crystal Fibers and Thin Rods Grown by Laser Heated Pedestal Growth

Directory of Open Access Journals (Sweden)

Gisele Maxwell

2017-01-01

Full Text Available Single-crystal fibers are an intermediate between laser crystals and doped glass fibers. They have the advantages of both guiding laser light and matching the efficiencies found in bulk crystals, which is making them ideal candidates for high-power laser and fiber laser applications. This work focuses on the growth of a flexible fiber with a core of dopant (Er, Nd, Yb, etc. and a polycrystalline clad of yttrium aluminum garnet (YAG that will exhibit good wave guiding properties. Direct growth or a combination of growth and cladding experiments are described. Scattering loss measurements at visible wavelengths, along with dopant profile characterization with damage threshold results, are also presented. For single-pass amplification, a single-pass linear gain of 7.4 was obtained for 29 nJ pulses of 5 ns duration at 1 MHz repetition rate. We also obtained a laser efficiency of over 58% in a diode-pumped configuration. These results confirm the potential for single-crystal fibers to overcome the limitations of the glass fibers commonly used in fiber lasers, making them prime candidates for high-power compact fiber lasers and amplifiers.

Temperature measurement in cans of fuel rods and fuel rod simulators

International Nuclear Information System (INIS)

Tschoeke, H.; Moeller, R.

1977-01-01

On the sodium-cooled 19-rod cluster model for the SNR 300 the can wall temperature distributions of the non-uniformly cooled rods were measured with thermocouples mounted in outer grooves in the peripheral zone, permitting, in connection with Ni solder, a practically undisturbed measurement. For a more exact determination of the local surface temperature a calibration method, the so-called double-wall method, was developed and applied. The description of this calibration method and the experimental results achieved until now are presented. (orig./RW) [de

Numerical Simulations on the Laser Spot Welding of Zirconium Alloy Endplate for Nuclear Fuel Bundle Assembly

Science.gov (United States)

Satyanarayana, G.; Narayana, K. L.; Boggarapu, Nageswara Rao

2018-03-01

In the nuclear industry, a critical welding process is joining of an end plate to a fuel rod to form a fuel bundle. Literature on zirconium welding in such a critical operation is limited. A CFD model is developed and performed for the three-dimensional non-linear thermo-fluid analysis incorporating buoyancy and Marnangoni stress and specifying temperature dependent properties to predict weld geometry and temperature field in and around the melt pool of laser spot during welding of a zirconium alloy E110 endplate with a fuel rod. Using this method, it is possible to estimate the weld pool dimensions for the specified laser power and laser-on-time. The temperature profiles will estimate the HAZ and microstructure. The adequacy of generic nature of the model is validated with existing experimental data.

Design of a high-power, high-brightness Nd:YAG solar laser.

Science.gov (United States)

Liang, Dawei; Almeida, Joana; Garcia, Dário

2014-03-20

A simple high-power, high-brightness Nd:YAG solar laser pumping approach is presented in this paper. The incoming solar radiation is both collected and concentrated by four Fresnel lenses and redirected toward a Nd:YAG laser head by four plane-folding mirrors. A fused-silica secondary concentrator is used to compress the highly concentrated solar radiation to a laser rod. Optimum pumping conditions and laser resonator parameters are found through ZEMAX and LASCAD numerical analysis. Solar laser power of 96 W is numerically calculated, corresponding to the collection efficiency of 24  W/m². A record-high solar laser beam brightness figure of merit of 9.6 W is numerically achieved.

Development of cutting device for irradiated fuel rod

International Nuclear Information System (INIS)

Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

2004-01-01

Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

Fabrication Of Control Rod System Of The RSG-GAS

International Nuclear Information System (INIS)

Sudirdjo, Hari; Setyono; Prasetya, Hendra

2001-01-01

Eight units of control rod mechanical system of RSG-GAS has been fabricated. The control rod mechanical system of RSG-GAS consist of guide tube and lifting rod. Complete construction of the control rod mechanical system of RSG-GAS are guide tube, lifting rod, absorber, and absorber casing. The eight units of the control rod mechanical system of RSG-GAS has been fabricated according to the mechanical engineering design

Microcomputer system for controlling fuel rod length

International Nuclear Information System (INIS)

Meyer, E.R.; Bouldin, D.W.; Bolfing, B.J.

1979-01-01

A system is being developed at the Oak Ridge National Laboratory (ORNL) to automatically measure and control the length of fuel rods for use in a high temperature gas-cooled reactor (HTGR). The system utilizes an LSI-11 microcomputer for monitoring fuel rod length and for adjusting the primary factor affecting length. Preliminary results indicate that the automated system can maintain fuel rod length within the specified limits of 1.940 +- 0.040 in. This system provides quality control documentation and eliminates the dependence of the current fuel rod molding process on manual length control. In addition, the microcomputer system is compatible with planned efforts to extend control to fuel rod fissile and fertile material contents

A LMFBR for thorium utilization and for the U233/Th fuel rods specification

International Nuclear Information System (INIS)

Ishiguro, Y.; Dias, A.F.

1982-01-01

The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

International Nuclear Information System (INIS)

Moriwaki, Masanao; Aoyama, Motoo; Anegawa, Takafumi; Okada, Hiroyuki; Sakurada, Koichi; Tanabe, Akira

2004-01-01

An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor

Feasibility study of a CO2-laser based lightning-protection system realization

Science.gov (United States)