Code Analyses Supporting PIE of Weapons-Grade MOX Fuel

International Nuclear Information System (INIS)

Ott, Larry J.; Bevard, Bruce Balkcom; Spellman, Donald J.; McCoy, Kevin

2010-01-01

The U.S. Department of energy has decided to dispose of a portion of the nation's surplus weapons-grade plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating the fuel in commercial power reactors. Four lead test assemblies (LTAs) were manufactured with weapons-grade mixed oxide (WG-MOX) fuel and irradiated in the Catawba Nuclear Station Unit 1, to a maximum fuel rod burnup of ∼47.3 GWd/MTHM. As part of the fuel qualification process, five rods with varying burnups and initial plutonium contents were selected from one assembly and shipped to the Oak Ridge National Laboratory (ORNL) for hot cell examination. ORNL has provided analytical support for the post-irradiation examination (PIE) of these rods via extensive fuel performance modeling which has aided in instrument settings and PIE data interpretation. The results of these fuel performance simulations are compared in this paper with available PIE data.

Program description for the qualification of CNEA - Argentina as a supplier of LEU silicide fuel and post-irradiation examinations plan for the first prototype irradiated in Argentina

International Nuclear Information System (INIS)

Rugirello, Gabriel; Adelfang, Pablo; Denis, Alicia; Zawerucha, Andres; Marco, Agustin di; Guillaume, Eduardo; Sbaffoni, Monica; Lacoste, Pablo

1998-01-01

In this report we present a description of the ongoing and future stages of the program for the qualification of CNEA, Argentina, as a supplier of low enriched uranium silicide fuel elements for research reactor. Particularly we will focus on the characteristics of the future irradiation experiment on a new detachable prototype, the post-irradiation examinations (PIE) plan for the already irradiated prototype PO4 and an overview of the recently implemented PIE facilities and equipment. The program is divided in several steps, some of which have been already completed. It concludes: development of the uranium silicide fissile material, irradiation and PIE of several full-scale prototypes. Important investments have been already carried out in the facilities for the FE production and PIE. (author)

Available post-irradiation examination techniques at Romanian institute for nuclear research

International Nuclear Information System (INIS)

Parvan, Marcel; Sorescu, Antonius; Mincu, Marin; Uta, Octavian; Dobrin, Relu

2005-01-01

The Romanian Institute for Nuclear Research (INR) has a set of nuclear facilities consisting of TRIGA 14 MW(th) materials testing reactor and LEPI (Romanian acronym for post-irradiation examination laboratory) which enable to investigate the behaviour of the nuclear fuel and materials under various irradiation conditions. The available techniques of post-irradiation examination (PIE) and purposes of PIE for CANDU reactor fuel are as follows. 1) Visual inspection and photography by periscope: To examine the surface condition such as deposits, corrosion etc. 2) Eddy current testing: To verify the cladding integrity. 3) Profilometry and length measurement performed both before and after irradiation: To measure the parameters which highlight the dimensional changes i.e. diameter, length, diametral and axial sheath deformation, circumferential sheath ridging height, bow and ovality. 4) Gamma scanning and Tomography: To determine the burnup, axial and radial fission products activity distribution and to check for flux peaking and loading homogeneity. 5) Puncture test: To measure the pressure, volume and composition of fission gas and the inner free volume. 6) Optical microscopy: To highlight the structural changes and hydriding, to examine the condition of the fuel-sheath interface and to measure the oxide thickness and Vickers microhardness. 7) Mass spectrometry: To measure the burnup. 8) Tensile testing: To check the mechanical properties. So far, non-destructive and destructive post-irradiation examinations have been performed on a significant number of CANDU fuel rods (about 100) manufactured by INR and irradiated to different power histories in the INR 14 MW(th) TRIGA reactor. These examinations have been performed as part of the Romanian research programme for the manufacturing, development and safety of the CANDU fuel. The paper describes the PIE techniques and some results. (Author)

Post irradiation examination of control rod assembly of FBTR

International Nuclear Information System (INIS)

Anandaraj, V.; Raghu, N.; Venkiteswaran, C.N.; Visweswaran, P.; Vijayakumar, Ran; Jayaraj, V.V.; Padmaprabu, P.; Saravanan, T.; Philip, John; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.

2010-01-01

Six control rods with boron carbide pellets are used in FBTR for shutdown and control of reactor power. One control rod after being subjected to a fluence level of 7.2 x 10 22 n/cm 2 was received for post irradiation examination (PIE) to assess its irradiation behavior and to investigate the incident of dropping of control rod. Examinations carried out include precise dimensional measurements to investigate the possibility of interference between the control rod and outer sheath, Neutron radiography and x-radiograph to assess the integrity of the boron carbide pellets and other internals, density measurements to assess the swelling behaviour of boron carbide pellets and metallographic examinations to study the cracking behaviour and microstructural changes in the pellet and the clad. Depletion of B 10 in the pellet was studied using time of flight mass spectrometry. The paper highlights the examinations and results of the PIE carried out. (author)

PIE Report on the KOMO-3 Irradiation Test Fuels

International Nuclear Information System (INIS)

Park, Jong Man; Ryu, H. J.; Yang, J. H.

2009-04-01

In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

PIE on Safety-Tested AGR-1 Compact 5-1-1

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-08-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

Current status of PIE activities in O-arai Engineering Center of JNC on FBR MOX fuel

International Nuclear Information System (INIS)

Koyama, Shin-ichi; Osaka, Masahiko; Namekawa, Takashi; Itoh, Masahiko

2003-01-01

Japan Nuclear Cycle Development Institute (JNC) is now totally promoting the development of commercialized fast reactors to realize stable supply of energy in future. One of the important items is to develop high-performance fuel. For this purpose, it is essential to carry out post-irradiation examinations (PIE) for evaluation of irradiated fuel performance and also to establish the PIE technology. This paper describes the current status of PIE results including its technology in O-arai Engineering Center of JNC. The facilities have been operating safely and successfully since the 1960's. Obtained PIE data were reflected to the design and operation of the experimental fast reactor JOYO, the prototype fast reactor MONJU and future fast reactors. The core modification from the breeding core (MK-I) to the irradiation core (MK-II) of JOYO was performed in 1982. Irradiation tests of fuels and materials in MK-II core started in 1982. At PIE facilities in OEC, 65 of driver fuels, fuel irradiation test rigs, material irradiation test rigs and several other components were examined related to JOYO MK-II core operation, and thus a lot of aspects were accumulated for irradiated fuel behaviors. As topical activities of these PIE techniques, burnup measurement and analytical technique for Minor Actinides (MA), such as neptunium and americium were described here. (author)

PIE analysis for minor actinide

International Nuclear Information System (INIS)

Suyama, Kenya

2005-01-01

Minor actinide (MA) is generated in nuclear fuel during the operation of power reactor. For fuel design, reactivity decrease due to it should be considered. Out of reactors, MA plays key role to define the property of spent fuel (SF) such as α-radioactivity, neutron emission rate, and criticality of SF. In order to evaluate the calculation codes and libraries for predicting the amount of MA, comparison between calculation results and experimentally obtained data has been conducted. In this report, we will present the status of PIE data of MA taken by post irradiation examinations (PIE) and several calculation results. (author)

Post-irradiation examination of overheated fuel bundles

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Leach, D.A.

1995-01-01

Post-irradiation examinations (PIE) were conducted on prototype 43-element CANDU fuel bundles that overheated during test irradiations in the NRU reactor. PIE revealed that the bundles remained physically intact, but on several elements the Zr-4 sheath collapsed into axial gaps between the pellet stack and end caps, between adjacent pellets within the stacks, and into missing pellet chips and cracks. Helium pressurization tests showed that none of the collapsed elements leaked. Hydride blisters were discovered on a few elements, but the source of the hydrogen was not linked to a breach of the cladding or end caps. These defects were attributed to primary hydriding. Microstructural changes in the fuel and cladding indicate that the cladding-was briefly exposed to temperatures in the range 600-800 o C and pressures above 11.2 MPa. The results show that Zr-4 cladding behaves in a highly ductile manner during such transient, high-temperature and high-pressure excursions. (author)

Post-irradiation examination of overheated fuel bundles

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Leach, D.A.

1997-08-01

Post-irradiation examinations (PIE) were conducted on prototype 43-element CANDU fuel bundles that overheated during test irradiations in the NRU reactor. PIE revealed that the bundles remained physically intact, but on several elements the Zr-4 sheath collapsed into axial gaps between the pellet stack and end caps, between adjacent pellets within the stacks, and into missing pellet chips and cracks. Helium pressurization tests showed that none of the collapsed elements leaked. Hydride blisters were discovered on a few elements, but the source of the hydrogen was.not linked to a breach of the cladding or end caps. These defects were attributed to primary hydriding. Microstructural changes in the fuel and cladding indicate that the cladding was briefly exposed to temperatures in the range 600-800 o C and pressures above 11.2MPa. The results show that Zr-4 cladding behaves in a highly ductile manner during such transient, high-temperature and high-pressure excursions. (author)

State of the VVER-1000 spent U-Gd fuel rods based on the results of post-irradiation examinations

International Nuclear Information System (INIS)

Shevlyakov, G.; Zvir, E.; Strozhuk, A.; Polenok, V.; Sidorenko, O.; Volkova, I.; Nikitin, O.

2015-01-01

The present paper is devoted to post-irradiation examinations (PIE) of U-Gd fuel rods with different geometry of the fuel pellets irradiated as part of the VVER-1000 fuel assembly. As evidenced by their PIE data, they did not exhaust their service life based on the main parameters (geometrical dimensions, corrosion state, and release of fission product gases). (author)

Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

International Nuclear Information System (INIS)

Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

1996-01-01

Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

Advanced post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2002-03-01

The purpose of the meeting was to provide and overview of the status of post-irradiation examination (PIE) techniques for water cooled reactor fuel assemblies and their components with emphasis given to advanced PIE techniques applied to high burnup fuel. Papers presented at the meeting described progress obtained in non-destructive (e.g. dimensional measurements, oxide layer thickness measurements, gamma scanning and tomography, neutron and X-ray radiography, etc.) and destructive PIE techniques (e.g. microstructural studies, elemental and isotopic analysis, measurement of physical and mechanical properties, etc.) used for investigation of water reactor fuel. Recent practice in high burnup fuel investigation revealed the importance of advanced PIE techniques, such as 3-D tomography, secondary ion mass spectrometry, laser flash, high resolution transmission and scanning electron microscopy, image analysis in microstructural studies, for understanding mechanisms of fuel behaviour under irradiation. Importance and needs for in-pile irradiation of samples and rodlets in instrumented rigs were also discussed. This TECDOC contains 20 individual papers presented at the meeting; each of the papers has been indexed separately

Post-irradiation examination of U3SIX-AL fuel element manufactured and irradiated in Argentina

International Nuclear Information System (INIS)

Ruggirello, Gabriel; Calabroni, Hector; Sanchez, Miguel; Hofman, Gerard

2002-01-01

As a part of CNEA's qualification program as a supplier of low enriched Al-U 3 Si 2 dispersion fuel elements for research reactors, a post irradiation examination (PIE) of the first prototype of this kind, called P-04, manufactured and irradiated in Argentina, was carried out. The main purpose of this work was to set up various standard PIE techniques in the hot cell, looking forward to the next steps of the qualification program, as well as to acquire experience on the behaviour of this nuclear material and on the control of the manufacturing process. After an appropriate cooling period, on May 2000 the P-04 was transported to the hot cell in Ezeiza Atomic Centre. Non destructive and destructive tests were performed following the PIE procedures developed in Argonne National Laboratory (ANL), this mainly included dimensional measurement, microstructural observations and chemical burn-up analyses. The methodology and results of which are outlined in this report. The results obtained show a behaviour consistent with that of other fuel elements of the same kind, tested previously. On the other hand the results of this PIE, specially those concerning burn-up analysis and stability and corrosion behaviour of the fuel plates, will be of use for the IAEA Regional Program on the characterization of MTR spent fuel. (author)

The MEGAPIE PIE sample preparation

International Nuclear Information System (INIS)

Wohlmuther, M.; Boutellier, V.; Dai, Y.; Gavillet, D.; Geissmann, K.; Hahl, S.; Hammer, B.; Lagotzki, A.; Leu, H.; Linder, H.P.; Kalt, A.; Kuster, D.; Neuhausen, J.; Schumann, D.; Schwarz, R.; Schweikert, H.; Spahr, A.; Suter, P.; Teichmann, S.; Thomsen, K.; Wiese, H.; Wagner, W.; Zimmermann, U.; Zumbach, C.

2015-01-01

On the way towards Accelerator-driven Systems (ADS), the MEGAPIE (Mega-Watt Pilot Experiment) project is one of the key milestones. The MEGAPIE project aimed to prove that a liquid Lead-Bismuth-Eutectic (LBE) spallation target can be licensed, planned, built, operated, dismantled, examined and disposed. The project has finished the phase of producing the samples for Post-irradiation Examination (PIE). Samples to study structural material property changes due to the harsh environment of high temperatures, contact with flowing liquid metal (LBE), proton and neutron irradiation will be investigated by all partner laboratories (CEA, CNRS, ENEA, KIT, PSI and SCK-CEN). (authors)

Hot cell facilities for post irradiation examination

International Nuclear Information System (INIS)

Mishra, Prerna; Bhandekar, Anil; Pandit, K.M.; Dhotre, M.P.; Rath, B.N.; Nagaraju, P.; Dubey, J.S.; Mallik, G.K.; Singh, J.L.

2017-01-01

Reliable performance of nuclear fuels and critical core components has a large bearing on the economics of nuclear power and radiation safety of plant operating personnel. In view of this, Post Irradiation Examination (PIE) is periodically carried out on fuels and components to generate feedback information which is used by the designers, fabricators and the reactor operators to bring about changes for improved performance of the fuel and components. Examination of the fuel bundles has to be carried out inside hot cells due to their high radioactivity

Recent and future PIE activities in KAERI

International Nuclear Information System (INIS)

Ahn, Sangbok; Ryu, Woo-seok; Hong, Kwonpyo; Jeon, Yongbum; Kim, Dosik; Choo, Yongsun

2012-01-01

KAERI has two PIE hot cell facilities. One is Irradiated Material Examination Facility (IMEF) to examine irradiated nuclear fuels and core structural materials at HANARO, and the other is a Post Irradiation Examination Facility (PIEF) to do the operated PWR nuclear fuels and skeleton. Their activities are categorized by hot cell examinations, new technique developments, maintenance of facility utilities and a public acceptance work. In the past three years hot cell tests were executed over 2,000 times related to R and D projects of the future and next generation reactors such as SMART, U-Mo fuels, I-NERI etc., including 30 rods to be operated in domestic PWR's. To supply PIE data on time to the user schedules new equipment and techniques are being developed for 10 items which include an annealing fission gas release rate, thermal diffusivity measurement, a laser puncturing fission gas collection system and mechanical property measurements on PWR FA skeleton etc. For maintenance of facilities the aged utility equipment are refurbished and exchanged for new ones. The partly cracked and leaking building walls are completely repaired, and the aged electric UPS's, HVAC system parts, heating and cooling system are exchanged for new ones. Additionally efforts to upgrade the public acceptance for facility visiting events, open training courses to the university and graduated students have been made. Regarding the future of domestic R and D and a power reactor operating plan the demand for hot cell examinations will be steadily increased. Along with them new PIE techniques on TRISO fuels, minor actinide SFR fuels and CANDU fuels etc are required to be developed. New space to install small scale hot cells, and develop for the extreme conditioned test to meet future demands will be taken. (author)

Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

International Nuclear Information System (INIS)

Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

2009-01-01

The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

Post irradiation examinations on HTTR materials

International Nuclear Information System (INIS)

Sakai, Haruyuki; Ohmi, Masao; Eto, Motokuni; Watanabe, Katsutoshi

1995-01-01

The HTTR (High Temperature engineering Test Reactor) is being constructed at Oarai Research Establishment of the Japan Atomic Energy Research Institute. In order to develop necessary materials for the HTTR, after irradiations in the JMTR, PIEs are being carried out on these materials in the JMTRHL (JMTR Hot Laboratory). Impact test, tensile test, fatigue test, creep test, metallography and so on were performed for irradiated 2 1/4Cr 1Mo steel as the pressure vessel material and Alloy 800H as the cladding material of the control rod. A fatigue testing machine and four creep testing machines newly designed were fabricated and installed in the steel cells in order to evaluate the integrity of the HTTR materials. The development process and PIE results obtained with these machines are given in this paper

Operation of post-irradiation examination facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Eun Ka; Park, Kwang Joon; Jeon, Yong Bum [and others; Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-02-01

In 1995, the post-irradiation examination (PIE) of nuclear fuels was performed as follows. The relation between burnup and top nozzle spring force of fuel assembly was obtained by measuring the holddown spring force on the Kori-1 reactor fuel assemblies. The resonance ultrasonic test for inspection of defect and moisture in fuel rod was carried out on fuel rods of C15 and J14 assemblies, and the change of fuel rod condition by storing in pool has been analyzed on the intentionally defected fuel rods (ID-C and ID-L) as well as intact fuel rod (1-2) by NDT in ht cell. The oxide layer thickness on cladding surface of J44-L12 fuel rod was measured by NDT method and metallography to reveal the oxidation as a function of temperature in the fuel rod, and the burnup of J44 fuel assembly was measured by chemical analysis. HVAC system and pool water treatment system of the PIE facility were continuously operated for air filtration and water purification. The monitoring of radiation and pool water in PIE facility has been carried out to maintain the facility safety, and electric power supply system was checked and maintained to supply the electric power to the facility normally. The developed measurement techniques of oxide layer thickness on fuel rod cladding and holddown spring force of top nozzle in fuel assembly were applied to examine the nuclear fuels. Besides, a radiation shielding glove box was designed and a hot cell compressor for volume reduction of radioactive materials was fabricated. 19 tabs., 38 figs., 7 refs. (Author) .new.

PIE of test assembly of Qinshan nuclear power plant

International Nuclear Information System (INIS)

Ran, M.; Yan, J.; Wang, S.

2000-01-01

The small dimensional test fuel assembly (3x3-2) for the Qinshan Nuclear Power Plant was irradiated up to 25.7 Gwd/tU in the in-pile loop (15.5 Mpa,320 C) in Heavy Water Research Reactor (HWRR), CIAE, at simulative condition to Qinshan PWR normal and short time overpower operation for verifying the design, technology, and material properties of the fuel assembly. Comprehensive post-irradiation examination (PIE) including dimension measurement, gamma scanning, eddy current test, X ray, radiography, measurement of fission gas release, and quantitative metallography etc. were performed. PIE results show that the diameter of the fuel rods changed, ridges appeared on the cladding, pellets swelled, and the rate of fission gas release was higher than what we expected. The results would be an important basis for further improvement of design, technology and material properties for Qinshan PWR assembly. (author)

Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

International Nuclear Information System (INIS)

Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

1985-01-01

Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e. fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations

Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

Energy Technology Data Exchange (ETDEWEB)

Reitsamer, G; Proksch, E; Stolba, G; Strigl, A; Falta, G; Zeger, J [Department of Chemistry, Austrian Research Centre Seibersdorf Ltd., Seibersdorf (Austria)

1985-07-01

Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e., fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations.

Post-irradiation examination of CANDU fuel bundles fuelled with (Th, Pu)O2

International Nuclear Information System (INIS)

Karam, M.; Dimayuga, F.C.; Montin, J.

2010-01-01

AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O 2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt.% Th and 1.53 wt.% Pu in (Th, Pu)O 2 . The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O 2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O 2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O 2 fuel performance characteristics were superior to UO 2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

ORNL capability to conduct post irradiation examination of full-length commercial nuclear fuel rods

International Nuclear Information System (INIS)

Spellman, Donald J.

2007-01-01

Hot cells at Oak Ridge National Laboratory (ORNL) are nearing completion of a multi-year upgrade program to implement 21. century capabilities to meet the examination demands for higher burnup fuels and the future demands that will come from fuel recycling programs. Fuel reliability and zero tolerance for fuel failure is more than an industry goal. Fuel reliability is becoming a requirement that supports the renaissance of nuclear power generation. Thus, fuel development and management of new forms of waste that will come from programs such as the Global Nuclear Energy Partnership (GNEP) will require extensive use of the flexible, high-quality, technically advanced hot cells at ORNL. ORNL has the capability to perform post irradiation examination (PIE) of irradiated commercial nuclear fuel rods and the management structure to ensure a timely, cost-effective result. ORNL can: 1) Handle the transportation issues, 2) Perform macroscopic fuel rod examinations, 3) Perform microscopic fuel and clad examinations, and 4) Manage legacy material and waste disposal issues from PIE activities. All four of these items will be managed in a way that allows the customer day-to-day access to the results and data. Hot cell examination equipment that is necessary to determine the characteristics and performance of irradiated materials must operate in a hostile environment and is subject to long-term degradation that may result in reliability and quality assurance (QA) issues. ORNL has modernized its hot cell nuclear fuel examination equipment, installing state-of-the-art automated examination equipment and data gathering capabilities. ORNL is planning a major commitment to nuclear fuel examination and development, and future improvements will continue to be made over the next few years. (author)

N Reactor pressure tube 1350 postirradiation examination

International Nuclear Information System (INIS)

Cook, D.J.

1977-01-01

The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

INL Initial Input to the Mission Need for Advanced Post-Irradiation Examination Capability A Non-Major System Acquisition Project

International Nuclear Information System (INIS)

Tonc, Vince

2010-01-01

Consolidated and comprehensive post-irradiation examination (PIE) capabilities will enable the science and engineering understanding needed to develop the innovative nuclear fuels and materials that are critical to the success of the U.S. Department of Energy's (DOE) Office of Nuclear Energy (NE) programs. Existing PIE capabilities at DOE Laboratories, universities, and in the private sector are widely distributed, largely antiquated, and insufficient to support the long-range mission needs. In addition, DOE's aging nuclear infrastructure was not designed to accommodate modern, state-of-the-art equipment and instrumentation. Currently, the U.S. does not have the capability to make use of state-of-the-art technology in a remote, hot cell environment to characterize irradiated fuels and materials on the micro, nano, and atomic scale. This 'advanced PIE capability' to make use of state-of-the-art scientific instruments in a consolidated nuclear operating environment will enable comprehensive characterization and investigation that is essential for effectively implementing the nuclear fuels and materials development programs in support of achieving the U.S. DOE-NE Mission.

KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

2016-11-01

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

Post Irradiation Capabilities at the Idaho National Laboratory

International Nuclear Information System (INIS)

Schulthess, J.L.; Rosenberg, K.E.

2011-01-01

The U.S. Department of Energy (DOE), Office of Nuclear Energy (NE) oversees the efforts to ensure nuclear energy remains a viable option for the United States. A significant portion of these efforts are related to post-irradiation examinations (PIE) of highly activated fuel and materials that are subject to the extreme environment inside a nuclear reactor. As the lead national laboratory, Idaho National Laboratory (INL) has a rich history, experience, workforce and capabilities for performing PIE. However, new advances in tools and techniques for performing PIE now enable understanding the performance of fuels and materials at the nano-scale and smaller level. Examination at this level is critical since this is the scale at which irradiation damage occurs. The INL is on course to adopt these advanced tools and techniques to develop a comprehensive nuclear fuels and materials characterization capability that is unique in the world. Because INL has extensive PIE capabilities currently in place, a strong foundation exist to build upon as new capabilities are implemented and work load increases. In the recent past, INL has adopted significant capability to perform advanced PIE characterization. Looking forward, INL is planning for the addition of two facilities that will be built to meet the stringent demands of advanced tools and techniques for highly activated fuels and materials characterization. Dubbed the Irradiated Materials Characterization Laboratory (IMCL) and Advanced Post Irradiation Examination Capability, these facilities are next generation PIE laboratories designed to perform the work of PIE that cannot be performed in current DOE facilities. In addition to physical capabilities, INL has recently added two significant contributors to the Advanced Test Reactor-National Scientific User Facility (ATR-NSUF), Oak Ridge National Laboratory and University of California, Berkeley.

Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

2017-08-15

The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

Directory of Open Access Journals (Sweden)

Jong Man Park

2017-08-01

Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

First results of the post-irradiation examination of the Ceramic Breeder materials from the Pebble Bed Assemblies Irradiation for the HCPB Blanket concept

International Nuclear Information System (INIS)

Hegeman, J.; Magielsen, A.J.; Peeters, M.; Stijkel, M.P.; Fokkens, J.H.; Laan, J.G. van der

2006-01-01

In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li 4 SiO 4 pebble-beds are operated at different maximum temperatures of approximately 600 o C and 800 o C. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li 2 TiO 3 beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li 2 TiO 3 was 800 o C. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

Pie technique of LWR fuel cladding fracture toughness test

International Nuclear Information System (INIS)

Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu

2006-01-01

Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)

EDF requirements for hot cells examinations on irradiated fuel

International Nuclear Information System (INIS)

Segura, J.C.; Ducros, G.

2002-01-01

The objectives of increasing French Nuclear Power Plants (NPP) availability while lengthening the fuel irradiation cycle and reaching higher burnups lead EDF to carry out on site and hot cell examinations. The data issued from such fuel behaviour monitoring programmes will be used to ascertain that the design criteria are met. Data are also needed for modelling, development and validation. The paper deals quickly with the logistics linked to the selection and transport of fuel rods from NPP to hot cell laboratory. Hot cell PIEs remain a valuable method to obtain data in such fields as PCI (Pellet-Cladding Interaction), internal pressure, FGR (Fission Gas Release), oxide thickness, metallurgical aspects. The paper introduces burnup determination methods, inner pressure evaluation, preparation of samples for further irradiation such as power ramps for PCI and RIA (Reactivity Initiated Accident) testing. The nuclear microprobe of Perre Suee laboratory is also presented. (author)

Operation of post-irradiation examination facility

Energy Technology Data Exchange (ETDEWEB)

Kim, E. G.; Jeon, Y. B.; Ku, D. S.

1996-12-01

In 1996, the post-irradiation examination(PIE) of nuclear fuels was performed as follows. It has been searched for the caution of defection of defected fuel rods of Youngkwang-4 reactor through NDT and metallographic examination that had been required by KEPCO. And in-pool inspection of Kori-1 spent fuel assembly(FO2) was carried out. HVAC system and pool water treatment system have been operated to maintain the facility safely, and electric power supply system was checked and maintained for the normal and steady supply electric power to the facility. Image processing software was developed for measurement of defection of spent fuel rods. Besides, a radiation shielding glove box was fabricated and a hot cell compressor for volume reduction of radioactive materials was fabricated and installed in hot cell. Safeguards of nuclear materials were implemented in strict accordance with the relevant Korean rules and regulations as well as the international non-proliferation regime. Also the IAEA inspection was carried out on the quarterly basis. (author). 31 tabs., 71 figs., 4 refs.

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

The operation of post-irradiation examination facility

International Nuclear Information System (INIS)

Kim, Eun Ka; Min, Duk Ki; Lee, Young Kil

1994-12-01

The operation of post-irradiation examination facility was performed as follow. HVAC and pool water treatment system were continuously operated, and radiation monitoring in PIE facility has been carried out to maintain the facility safely. Inspection of the fuel assembly (F02) transported from Kori Unit 1 was performed in pool, and fuel rods extracted from the fuel assembly (J44) of Kori Unit 2 NPP were examined in hot cell. A part of deteriorated pipe line of drinking water was exchanged for stainless steel pipe to prevent leaking accidents. Halon gas system was also installed in the exhausting blower room for fire fighting. And IAEA inspection camera for safeguard of nuclear materials was fixed at the wall in pool area. Radiation monitoring system were improved to display the area radioactive value at CRT monitor in health physics control room. And automatic check system for battery and emergency diesel generator was developed to measure the voltage and current of them. The performance test of oxide thickness measuring device installed in hot cell for irradiated fuel rod and improvement of the device were performed, and good measuring results using standard sample were obtained. The safeguard inspection of nuclear materials and operation inspection of the facility were carried out through the annual operation inspection, quarterly IAEA inspection and quality assurance auditing. 26 tabs., 43 figs., 14 refs. (Author) .new

The operation of post-irradiation examination facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Eun Ka; Park, Kwang Jun; Lee, Won Sang [and others; Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-01-01

The operation and management of PIE facility was executed in 1993. An indigenous 16 x 16 PWR type fuel assembly (ID No. J44) which was discharged from Kori unit 2 power reactor was transported to KAERI`s PIE facility and in-pool nondestructive examination and hot cell examination for the fuel were carried out. Because the above-mentioned 16 x 16 fuel is different from 14 x 14 fuel in its size and array of fuel rods, several examination and handling equipment for the 16 x 16 type fuel were designed and fabricated. PIE facility was operated in normal condition with the periodical check and inspection of the devices. The filter housing on the roof hood in chemical analysis hot cell was modified mounting air pressure gauge to indicate the optimal filter exchanging time. The burst air heating coil plate and the broken blowing fan of the HVAC system were repaired. The defaced grand packing in pool water circulation pump was replaced with the mechanical seal to prevent the leakage from the pump shaft sealing. The radiation monitoring in the facility was carried out to maintain the safe working condition and several radiation monitors were repaired. Spare parts for the radiation monitoring system were prepared to maintain the facility safely. The performance test of the emergency electric power supply system including UPS, battery and diesel generator was carried out. Oxide layer thickness measuring device for the performance test. Several devices including spent fuel handling equipment for the 17 x 17 PWR type fuel assembly were designed and fabricated for the subsequent PIE of nuclear fuels. 35 tabs., 17 figs., 7 refs. (Author) .new.

Hot Cell Post-Irradiation Examination and Poolside Inspection of Nuclear Fuel. Proceedings of the IAEA-HOTLAB Technical Meeting

International Nuclear Information System (INIS)

2013-04-01

The growing operational requirements for nuclear fuel, such as longer fuel cycles, higher burnups and wider use of transient regimes, require more robust fuel designs and more radiation resistant materials. Development of such advanced fuels is only possible with testing and analysis of their performance and application of adequate post-irradiation examination (PIE) methods and techniques. In addition, operational feedback data from poolside and PIE facilities are absolutely necessary for verification of fuel modelling codes and analysis of fuel failure mechanisms. For these reasons, the International Atomic Energy Agency (IAEA) has supported the international exchange of knowledge and sharing of best practices in the application of modern destructive and non-destructive methods of investigation of highly radioactive materials through a series of technical meetings (TMs), the last of which was held in 2006 in Buenos Aires. Since 1963, similar meetings, initially at the European level, have been organized by the Hot Laboratories and Remote Handling Working Group (HOTLAB), a partner in the development of the IAEA's Post Irradiation Examination Facilities Database (PIEDB), part of the IAEA's Integrated Nuclear Fuel Cycle Information System. With this successful partnership in mind, in 2010 the IAEA Technical Working Group on Fuel Performance and Technology recommended that a joint IAEA-HOTLAB TM be held on 'Hot Cell Post-Irradiation Examination and Pool-Side Inspection of Nuclear Fuel', covering questions relevant to the IAEA sub-programmes on 'Nuclear Power Reactor Fuel Engineering' and 'Management of Spent Fuel from Nuclear Power Reactors'. The TM was held on 23-27 May 2011, in Smolenice, Slovakia, with the participation of a large number of interested organizations and comprehensive coverage of major PIE and poolside inspection issues relating to both operation and storage of fuel for nuclear power reactors. The proceedings, summaries and conclusions of that joint

Investigation on cause of outage of Wide Range Monitor (WRM) in High Temperature engineering Test Reactor (HTTR). Post Irradiation Examination (PIE) toward investigation of the cause

International Nuclear Information System (INIS)

Shinohara, Masanori; Saito, Kenji; Takada, Shoji; Ishimi, Akihiro; Katsuyama, Kozo; Motegi, Toshihiro

2012-08-01

An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the outage by specifying the damaged part causing the event in the WRM. The one is a post irradiation examination using the X-ray computed tomography scanner in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. The other is an experiment using a mock-up simulating the WRM fabricated by the fabricator. The characteristic impedance of the damaged WRM was measured by Time Domain Reflectometry, which was compared with that of the mock-up, which could narrow down the damaged part in the WRM. This report summarized the results of the PIE and the experimental investigation using the mock-up to reveal the cause of outage of WRM. (author)

Microstructure and elemental distribution of americium containing MOX fuel under the short term irradiation tests

International Nuclear Information System (INIS)

Tanaka, Kosuke; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin Ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

2008-01-01

In order to investigate the effect of americium addition to MOX fuels on the irradiation behavior, the 'Am-1' program is being conducted in JAEA. The Am-1 program consists of two short term irradiation tests of 10-minute and 24 hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. The PIEs for Am-containing MOX fuels focused on the microstructural evolution and redistribution behavior of Am at the initial stage of irradiation and the results to date are reported

Status report on the irradiation testing and post-irradiation examination of low-enriched U3O8-Al and UAlsub(x)-Al fuel element by the Netherlands Energy Research Foundation (ECN)

International Nuclear Information System (INIS)

Pruimboom, H.; Lijbrink, E.; Otterdijk, K. von; Swanenburg de Veye, R.J.

1984-01-01

Within the framework of the RERTR-programme four low-enriched (20%) MTR-type fuel elements have been irradiated in the High Flux Reactor at Petten (The Netherlands) and are presently subjected to postirradiation examination. Two of the elements contain UAlsub(x)-Al and two contain U 3 O 8 -Al fuel. The test irradiation has been completed up to the target burn-up values of 50% and 75% respectively. An extensive surveillance programme carried out during the test period has confirmed the excellent in-reactor behaviour of both types. Post-irradiation examination of the 50% burn-up test elements, comprising of dimensional measurements, burn-up determination, fuel metallography and blister testing, has sofar confirmed the irradiation experiences. Good agreement between calculated and measured power and burn-up characteristics has been found. A survey of the test element characteristics, their irradiation history, the irradiation tests and the preliminary PIE results is given in the paper. (author)

Post irradiation examination of tight fit garter springs from PHWR after 15 HOY of operation

International Nuclear Information System (INIS)

Dubey, J.S.; Shriwastaw, R.S.; Kumar, Ashwini; Shah, Priti Kotak; Rath, B.N.; Kumar, Sunil; Mishra, Prerna; Alur, V.D.; Mallik, G.K.; Anantharaman, S.

2015-01-01

Tight fit irradiated garter springs made of Zr-2.5Nb-0.5Cu, belonging to Pressurized Heavy Water Reactors (PHWRs), which had experienced 15 Hot Operation Years (HOY) of operation were subjected to visual, dimensional, metallographic examination and relevant mechanical tests such as stretch, crush and rolling-fatigue tests. The girdle wires, of these garter springs, made of Zr-2 were subjected to tension and bend tests. Methodology of the tests conducted and results of PIE are presented. Significance of the test results are discussed in view of its continued operation in reactor. (author)

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).

WWER fuel: Results of post irradiation examination

International Nuclear Information System (INIS)

Markov, D.V.; Smirnov, V.P.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

2006-01-01

Experience in the field of fabrication, operation, testing and post-irradiation examinations (PIE) made it possible to settle the following requirements for a new generation of WWER nuclear fuel: - For WWER-1000 FA, the service life is no less than 5 years, 3 alternative fuel cycles (FC): 12 months x 4 FCs, 12 months x 5 FCs and 18 months x 3 FCs; - For WWER-440 FA, fuel cycle is 12 months x 5 FCs and a part of operating assembly is left for the 6. year; - High fuel burnup - up to 70 MWd/kgU; - Dimensional stability of FA and its components; - FA repairability; - Adaptability of fuel cycles; - Maintenance of maneuvering operating conditions at the NPP; - Reliability of control rod operation; - High serviceability level - FE leakage is no worse than 10-5 l/year. In order to provide the fulfillment of the above-given requirements, designers and production engineers have worked out cumulative measures and engineering solutions, which are introduced in development of a new generation fuel. Currently old design FA-M assemblies provided with steel skeleton are being operated in WWER-1000 reactors at Ukrainian and Bulgarian NPPs. As for Russian NPPs, new-type FAs are operated. These are advanced FAs (AFA), FA-A and FA-2 provided with zirconium alloy skeletons. A design of the second generation of WWER-440 operating assemblies was developed with respect to changes in some geometrical parameters, fastening of FEs in the lower grid (splinting was substituted for collet), usage of reinforcing rib under the lower grid, anti-debris filter and hafnium elements of junction unit as well as hafnium content decrease from 0.05 % mass down to 0.01% mass in zirconium materials. They are basic designs of FAs in order to be introduced in a five-year fuel cycle of WWER-440 NPPs in Czech Republic and Slovakia since 2005 and have got prospects for development. The operating experience of dismountable operating assemblies at the Loviisa NPP, vibration-proof operating assemblies at the

PIE and separate effect test of high burnup UO2 fuel

International Nuclear Information System (INIS)

Yang, Yong Sik; Kim, S.K.; Kim, D.H.

2005-01-01

To investigate the performance of a high burnup UO 2 fuel, the highest burnup fuel assembly in KOREA was transported to the PIE facility in KAERI. It was a 17·17 fuel assembly irradiated at the Ulchin Unit 2 PWR. The peak fuel rod average burnup was about 57MWd/kgU and locally 65MWd/kgU. The general PIE was performed to investigate the fuel rod irradiation performance. Fission gas release, burnup, oxide thickness, hydrogen pickup, CRUD, and density change were measured by destructive of non-destructive test. Microstructure change, bubble and pore size distributions were observed by optical microscopy, SEM and EPMA. All generated and available PIE results were used to verify high burnup fuel performance code INFRA. Several rods were cut for additional separate effect test. For the high burnup fission gas release behaviour analysis, annealing apparatus were developed and installed in hot cell and preliminary test was performed. In addition to current apparatus new induction furnace will be installed in hot cell to investigate the high temperature and transient fission gas release behaviour. Ring tensile test was performed to analyze the material property degradation which caused by the oxidation and hydride, and additional mechanical tests will be performed. (Author)

R and D status and requirements for PIE in the fields of the HTGR fuel and the innovative basic research on High-Temperature Engineering

Energy Technology Data Exchange (ETDEWEB)

Sawa, Kazuhiro; Tobita, Tsutomu; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Ishihara, Masahiro; Hayashi, Kimio; Hoshiya, Taiji; Sekino, Hajime; Ooeda, Etsurou

1999-09-01

The High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor (HTGR) in Japan, achieved its first criticality in November 1998 at the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI). In the field of HTGR fuel development, JAERI will proceed research and development (R and D) works by the following steps: (STEP-1) confirmation of irradiation performance of the first-loading fuel of the HTTR, (STEP-2) study on irradiation performance of high burnup SiC-coated fuel particle and (STEP-3) development of ZrC-coated fuel particle. Requirements for post-irradiation examination (PIE) are different for each R and D step. In STEP-1, firstly, hot cells will be prepared in the HTTR reactor building to handle spent fuels. In parallel, general equipments such as those for deconsolidation of fuel compacts and for handling coated fuel particles will be installed in the Hot Laboratory at Oarai. In STEP-2, precise PIE techniques, for example, Raman spectroscopy for measurement of stress on irradiated SiC layer, will be investigated. In STEP-3, new PIE techniques should be developed to investigate irradiation behavior of ZrC-coated particle. In the field of the innovative basic research on high-temperature engineering, some preliminary tests have been made on the research areas of (1) new materials development, (2) fusion technology, (3) radiation chemistry and (4) high-temperature in-core instrumentation. Requirements for PIE are under investigation, in particular in the field of the new materials development. Besides more general apparatuses including transmission electron microscopy (TEM), some special apparatuses such as an electron spin resonance (ESR) spectrometer, a specific resistance/Hall coefficient measuring system and a differential scanning calorimeter (DSC) are planned to install in the Hot Laboratory at Oarai. Acquisition of advanced knowledge on the irradiation behavior is expected in

Pie en espejo

OpenAIRE

Delgado Quiñones, Alberto; Pancorbo Sandoval, Enrique; Diaz Prieto, Giraldo; Martín Tirado, Juan Carlos; Hernández Varela, Duniesky; Cruz Alard, Ronald

2014-01-01

Paciente masculino de 79 años de edad, el cual presenta en el miembro inferior izquierdo a nivel del pie una malformación congénita que corresponde a un pie en espejo, representado por 9 artejos, articulados en 8 metatarsianos, pie zambo, seudoartrosis alta de la tibia y luxación congénita alta del peroné. Es inusual encontrar esta deformidad en el momento actual. Here is a 79 years-old patient who had in his left foot a congenital malformation called mirror foot, which is represented by 9...

PIE technology on mechanical tests for HTTR core component and structural materials developed at Research Hot Laboratory

International Nuclear Information System (INIS)

Kizaki, Minoru; Honda, Junichi; Usami, Kouji; Ouchi, Asao; Oeda, Etsuro; Matsumoto, Seiichiro

2001-02-01

The high temperature engineering test reactor (HTTR) with the target operation temperature of 950degC established the first criticality on November, 1998 based on a large amount of R and D results on fuel and materials. In such R and D works, the development of reactor materials are one of the key issues from the view point of reactor environments such as extremely high temperature, neutron irradiation and so on for the HTTR. The Research Hot Laboratory (RHL) had carried out much kind of post irradiation examinations (PIEs) on core component and pressure vessel materials for during more than a quarter century. And obtained data played an important role in development, characterization and licensing of those materials for the HTTR. This paper describes the PIE technology developed at RHL and typical results on mechanical tests such as elevated temperature tensile and creep rupture tests for Hasteloy-X, Incolloy 800H and so on, and Charpy impact, J IC fracture toughness, K Id fracture toughness and small punch tests for normalized and tempered 2 1/4Cr-1Mo steel from historical view. In addition, an electrochemical test technique established for investigating the irradiation embrittlement mechanism on 2 1/4Cr-1Mo steel is also mentioned. (author)

Development of crack growth and crack initiation test units for stress corrosion cracking examinations in high-temperature water environments under neutron irradiation (1) (Contract research)

International Nuclear Information System (INIS)

Izumo, Hironobu; Ishida, Takuya; Kawamata, Kazuo; Inoue, Shuichi; Ide, Hiroshi; Saito, Takashi; Ishitsuka, Etsuo; Chimi, Yasuhiro; Ise, Hideo; Miwa, Yukio; Ugachi, Hirokazu; Nakano, Junichi; Kaji, Yoshiyuki; Tsukada, Takashi

2009-04-01

To evaluate integrity of irradiation-assisted stress corrosion cracking (IASCC) on in-core structural materials used in light water reactors (LWRs), useful knowledge regarding IASCC has been obtained mainly by post-irradiation examinations (PIEs). In the core of commercial LWRs, however, the actual IASCC occurs under the effects of irradiation on both materials and high-temperature water environment. Therefore, it is necessary to confirm the suitability of the knowledge by PIE with comparison to IASCC behaviors during in-core SCC tests. Fundamental techniques for in-core crack growth and crack initiation tests have been developed already at the Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency (JAEA). For the in-core crack growth test technique, to evaluate the effects of neutron irradiation on stainless steels irradiated to low neutron fluences, it is indispensable to develop new loading technique which is applicable to compact tension (CT) specimens with thickness of 0.5 inch (0.5T), from the viewpoint of validity based on the fracture mechanics. Based on the present technical investigation for the in-core loading technique, it is expected that a target load of 7.6 kN approximately can apply to a 0.5T-CT specimen by adopting a loading unit of a lever type instead of the previous uni-axial tension type. For the in-core crack initiation test technique, moreover, construction of a loading unit adopting linear variable differential transformers (LVDTs) has been investigated and technical issues have examined. (author)

COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

OpenAIRE

ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

2013-01-01

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

International Nuclear Information System (INIS)

Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

1985-01-01

The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

Tratamiento del pie equino

OpenAIRE

Robles, Fernanda

2012-01-01

El pie equino varo es una patología frecuente en los pacientes con Encefalopatía crónica no evolutiva (ECNE).Se observó que existe una mayor frecuencia en el sexo masculino. Se realizó un estudio descriptivo, comparativo y secundario, durante el periodo de septiembre del 2006 a septiembre del 2011, con un total de 125 pacientes con Ecne, de los cuales se encontraron 101 con pie equino varo congénito. Objetivo: observar los resultados obtenidos en la rehabilitación del pie equino, en niños de ...

Microstructural evolution and Am migration behaviour in Am-containing fuels at the initial stage of irradiation

International Nuclear Information System (INIS)

Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Osaka, Masahiko; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin-ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

2010-01-01

In order to investigate the effect of americium addition to MOX fuels on the irradiation behaviour, the 'Am-1' programme is being conducted in JAEA. The Am-1 programme consists of two short-term irradiation tests of 10-minute and 24-hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post-irradiation examinations (PIE) are in progress. The PIE for Am-containing MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation and the results to date are reported. The successful development of fabrication technology with remote handling and the evaluation of thermo-chemical properties based on the out-of-pile experiments are described with an emphasis on the effects of Am addition on the MOX fuel properties. (authors)

Post-irradiation examinations and theoretical analyses of the fuel and fuel pin behavior of innovative systems

International Nuclear Information System (INIS)

Ernst, W.; Freund, D.; Geithoff, D.; Heck, M.; Jacobi, O.; Steiner, H.; Weimar, P.

1995-01-01

For the development of LMR cladding materials an irradiation test was performed in the KNK II reactor between 1986 and 1991. It demonstrated that a temperature gradient in the cladding has no observable effect on swelling behavior. PIEs of transient experiments in the Petten HFR have shown no significant cladding deformations. (orig.)

Irradiation test of component for radiation-resistant small sized motor

International Nuclear Information System (INIS)

Nakamichi, M.; Ishitsuka, E.; Shimakawa, S.; Kan, S.

2009-01-01

A small-sized motor with a resistance to radiation was developed. This motor has been able to operate at a gamma-ray dose of a value 700 times as high as the specification of a commercial motor. The present work describes results of post-irradiation examinations (PIEs) to evaluate effects of neutron irradiation on the lifetime of some major components of the motor such as a bearing, a magnet and a fixation agent for a field coil wire. It became clear from the results of PIEs that the radiation-resistance dose of the motor using a Sm-Co magnet will be expected to be one order of magnitude higher than that of the motor using a Nb-Fe-B magnet.

Pie de madura

Directory of Open Access Journals (Sweden)

Julio E. Torres Reyes

2014-08-01

Full Text Available En el Hospital provincial “Antonio Agosthino Neto” de la ciudad de NDalatando, Kwanza Norte, Republica de Angola, se recibe una paciente femenina de la raza negra de 20 años de edad, con antecedentes de trabajar en el campo, laborando habitualmente descalza. Desde hace dos años, después de haberse hecho una herida con una espina de acacia, presentaba lesiones en la región del tarso y metatarso del pie derecho. Las lesiones nodulares que abarcaban la planta y dorso del pie presentaban color rojo amarillento y supuraban constantemente, con aumento de partes blandas. Se le indicaron análisis complementarios y estudios imagenológicos. Después de obtener los resultados micológicos, bacteriológicos e imagenológicos y debido a las lesiones óseas y en partes blandas que presentaba la paciente, se decide la amputación quirúrgica del pie

El pie diabético

OpenAIRE

Cespedes Cespedes, Tomás; Cuevas Gómez, Rafael; Dorca Coll, Adelina; Sacristán Valero, Sergi

1991-01-01

El pie diabético está considerado como un pie de alto riesgo, está demostrado que si aplicáramos una adecuada y correcta educación sanitaria a todos los profesionales de la salud y al propio paciente, se podrían evitar en la mayoría de los casos las tan temidas amputaciones. En el presente trabajo hacemos referencia de las características más importantes de la diabetes, pie diabético y a la actuación profesional del podólogo frente a estos pacientes.

Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

International Nuclear Information System (INIS)

Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

2016-01-01

Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

National Postirradiation Examination Workshop Report

Energy Technology Data Exchange (ETDEWEB)

Schulthess, Jason L

2011-06-01

A National Post-Irradiation-Examination (PIE) Workshop was held March 29-30, 2011, in Washington D.C., stimulated by the DOE Acting Assistant Secretary for Nuclear Energy approval on January 31, 2011 of the “Mission Need Statement for Advanced Post-Irradiation Examination Capability”. As stated in the Mission Need, “A better understanding of nuclear fuels and material performance in the nuclear environment, at the nanoscale and lower, is critical to the development of innovative fuels and materials required for tomorrow’s nuclear energy systems.” (2011) Developing an advanced post-irradiation capability is the most important thing we can do to advance nuclear energy as an option to meeting national energy goals. Understanding the behavior of fuels and materials in a nuclear reactor irradiation environment is the limiting factor in nuclear plant safety, longevity, efficiency, and economics. The National PIE Workshop is part of fulfilling or addressing Department of Energy (DOE) missions in safe and publically acceptable nuclear energy. Several presentations were given during the opening of the workshop. Generally speaking, these presentations established that we cannot continue to rely on others in the world to provide the capabilities we need to move forward with nuclear energy technology. These presentations also generally identified the need for increased microstructural understanding of fuels and materials to be coupled with modeling and simulation, and increased accessibility and infrastructure to facilitate the interaction between national laboratories and participating organizations. The overall results of the work of the presenters and panels was distilled into four primary needs 1. Understanding material changes in the extreme nuclear environment at the nanoscale. Nanoscale studies have significant importance due to the mechanisms that cause materials to degrade, which actually occur on the nanoscale. 2. Enabling additional proficiency in

Experimental irradiation of UMo fuel: Pie results and modeling of fuel behaviour

International Nuclear Information System (INIS)

Languille, A.; Plancq, D.; Huet, F.; Guigon, B.; Lemoine, P.; Sacristan, P.; Hofman, G.; Snelgrove, J.; Rest, J.; Hayes, S.; Meyer, M.; Vacelet, H.; Leborgne, E.; Dassel, G.

2002-01-01

Seven full-sized U Mo plates containing ca. 8 g/cm 3 of uranium in the fuel meat have been irradiated since the beginning of the French U Mo development program. The first three of them with 20% 235 U enrichment were irradiated at maximum surfacic power under 150 W/cm 2 in the OSIRIS reactor up to 50% burn-up and are under examination. Their global behaviour is satisfactory: no failure and a low swelling. The other four plates were irradiated in the HFR Petten at maximum surfacic power between 150 and 250 W/cm 2 with two enrichments 20 and 35%. The experiment was stopped after two cycles due to a fuel failure. The post- irradiation examinations were completed in 2001 in Petten. Examinations showed a correct behaviour of 20% enriched plates and an abnormal behaviour of the two other plates (35%-enriched) with a clad failure on the plate 4. The fuel failure appears to result from a combination of factors that led to high corrosion cladding and high fuel meat temperatures. (author)

Post-Irradiation Examination and In-Pile Measurement Techniques for Water Reactor Fuels

International Nuclear Information System (INIS)

2009-12-01

in the 1960s when the construction of NPPs was being started. Evidently it can be assumed that infrastructure with basic unique equipments is old enough, both morally and physically, and needs to be up-graded or replaced. Thus, a sharp increase of the hydrocarbon fuel cost, green-house effect, necessity to construct more safe and efficient NPPs, justification of the lifetime prolongation of the existing NPPs, moral and physical ageing of the hot labs and research reactors equipment lead to the strong necessity to develop more perfect and more precise methods and equipment to examine irradiated components of nuclear reactors, first of all the most expensive one - nuclear fuel. Now the national hot laboratories and material testing reactors usually act as individual independent research establishments without any common and coordinated technical and business strategy towards the future needs and challenges. Even if there are not many joint programs for the development of nuclear power engineering in different countries, the method base and accumulated experience of the in- and post-reactor experiments should be widely shared so as to decrease the cost of this base in each country and to enforce its development. Thus, both problems and results of the application of new techniques to examine nuclear reactor components, as well as the conditions of separate labs should be discussed at the international level. The IAEA technical meetings are one of the most convenient means of arranging such discussion on the problems of the hot labs and research reactors development and application of new original techniques for examination of reactor materials properties. This publication represents a summary and proceedings of the two technical meetings (TMs) organized by IAEA on the subjects of Hot Cell Post-Irradiation Examination (PIE) Techniques and Pool Side Inspection of Water Reactor Fuel Assemblies and Fuel Rod Instrumentation and In-Pile Measurement Techniques. The first TM was

An investigation on the irradiation behavior of atomized U-Mo/Al dispersion rod fuels

International Nuclear Information System (INIS)

Park, J.M.; Ryu, H.J.; Lee, Y.S.; Lee, D.B.; Oh, S.J.; Yoo, B.O.; Jung, Y.H.; Sohn, D.S.; Kim, C.K.

2005-01-01

The second irradiation fuel experiment, KOMO-2, for the qualification test of atomized U-Mo dispersion rod fuels with U-loadings of 4-4.5 gU/cc at KAERI was finished after an irradiation up to 70 at% U 235 peak burn-up and subjected to the IMEF (Irradiation material Examination Facility) for a post-irradiation analysis in order to understand the fuel irradiation performance of the U-Mo dispersion fuel. Current results for PIE of KOMO-2 revealed that the U-Mo/Al dispersion fuel rods exhibited a sound performance without any break-away swelling, but most of the fuel rods irradiated at a high linear power showed an extensive formation of the interaction phase between the U-Mo particle and the Al matrix. In this paper, the analysis of the PIE results, which focused on the diffusion related microstructures obtained from the optical and EPMA (Electron Probe Micro Analysis) observations, will be presented in detail. And a thermal modeling will be carried out to calculate the temperature of the fuel rod during an irradiation. (author)

Evaluation of the characteristics of high burnup and high plutonium content mixed oxide (MOX) fuel

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-08-15

Two kinds of MOX fuel irradiation tests, i.e., MOX irradiation test up to high burnup and MOX having high plutonium content irradiation test, have been performed from JFY 2007 for five years in order to establish technical data concerning MOX fuel behavior during irradiation, which shall be needed in safety regulation of MOX fuel with high reliability. The high burnup MOX irradiation test consists of irradiation extension and post irradiation examination (PIE). The activities done in JFY 2011 are destructive post irradiation examination (D-PIE) such as EPMA and SIMS at CEA (Commissariat a l'Enegie Atomique) facility. Cadarache and PIE data analysis. In the frame of irradiation test of high plutonium content MOX fuel programme, MOX fuel rods with about 14wt % Pu content are being irradiated at BR-2 reactor and corresponding PIE is also being done at PIE facility (SCK/CEN: Studiecentrum voor Kernenergie/Centre d'Etude l'Energie Nucleaire) in Belgium. The activities done in JFY 2011 are non-destructive post irradiation examination (ND-PIE) and D-PIE and PIE data analysis. In this report the results of EPMA and SIMS with high burnup irradiation test and the result of gamma spectrometry measurement which can give FP gas release rate are reported. (author)

An automated tensile machine for small specimens heavily neutron irradiated in FFTF/MOTA

International Nuclear Information System (INIS)

Kohyama, Akira; Sato, Shinji; Hamada, Kenichi

1993-01-01

The objective of this work is to develop a fully automated tensile machine for post-irradiation examination (PIE) of Fast Flux Test Facility (FFTF)/Materials Open Test Assembly (MOTA) irradiated miniature tension specimens. The anticipated merit of the automated tensile machine is to reduce damage to specimens during specimen handling for PIE and to reduce exposure to radioactive specimens. This machine is designed for testing at elevated temperatures, up to 873 K, in a vacuum or in an inert gas environment. Twelve specimen assemblies are placed in the vacuum chamber that can be tested successively in a fully automated manner. A unique automated tensile machine for the PIE of FFTF/MOTA irradiated specimens, the Monbusho Automated Tensile Machine (MATRON) consists of a test frame with controlling units and an automated specimen-loading apparatus. The qualification of the test frame has been completed, and the results have satisfied the machine specifications. The capabilities of producing creep and relaxation data have been demonstrated for Cu, Al, 316SS, and ferritic steels. The specimen holders for the three-point bending test and the small bulge test (small punch test; SP test) were also designed and produced

Pie plano-valgo adquirido

OpenAIRE

Albiol Ferrer, Josep Maria; Giralt de Veciana, Enrique; Marugán de los Bueis, Montse; Novel Martí, Virginia; Ogalla, José Manuel; Zalacain, Antonio

1990-01-01

Es la patología del pie que tiene mayor incidencia durante la infancia. Su recuperación eficaz, además de precisar un largo período de tratamiento, requiere un diagnóstico precoz. Durante la lactancia, a partir de los tres meses, y después, durante la infancia, hasta los tres años aproximadamente, una gruesa y blanda capa de panlculo adiposo determina una configuración que rellena la bóveda y ofrece una falsa imagen de pie plano.

Nuestro concepto actual del pie valgo

OpenAIRE

Dorca Coll, Adelina; Cespedes Cespedes, Tomás; Concustell Gonfaus, Josep; Sacristán Valero, Sergi; Dorca Coll, Ma. Rosa

1996-01-01

Los autores hacen una revisión de los conceptos y etiologías que se han descrito del pie valgo. Incluyen una definición personal sobre el concepto funcional del pie valgo, definen por primera vez la marcha que cursa con estrés en valgo y presentan una descripción del comportamiento biomecánico de los diferentes pies valgos. Esta diferenciación será de gran utilidad para la posterior aplicación del tratamiento ortopodológico adecuado. Es de resaltar las imágenes que se presentan obtenidas de c...

9 CFR 319.500 - Meat pies.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Meat pies. 319.500 Section 319.500... ORGANIZATION AND TERMINOLOGY; MANDATORY MEAT AND POULTRY PRODUCTS INSPECTION AND VOLUNTARY INSPECTION AND CERTIFICATION DEFINITIONS AND STANDARDS OF IDENTITY OR COMPOSITION Meat Food Entree Products, Pies, and...

Neutron Focusing Mirrors for Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory

Science.gov (United States)

Rai, Durgesh K.; Wu, Huarui; Abir, Muhammad; Giglio, Jeffrey; Khaykovich, Boris

Post irradiation examination (PIE) of samples irradiated in nuclear reactors is a challenging but necessary task for the development on novel nuclear power reactors. Idaho National Laboratory (INL) has neutron radiography capabilities, which are especially useful for the PIE of irradiated nuclear fuel. These capabilities are limited due to the extremely high gamma-ray radiation from the irradiated fuel, which precludes the use of standard digital detectors, in turn limiting the ability to do tomography and driving the cost of the measurements. In addition, the small 250 kW Neutron Radiography Reactor (NRAD) provides a relatively weak neutron flux, which leads to low signal-to-noise ratio. In this work, we develop neutron focusing optics suitable for the installation at NRAD. The optics would separate the sample and the detector, potentially allowing for the use of digital radiography detectors, and would provide significant intensity enhancement as well. The optics consist of several coaxial nested Wolter mirrors and is suited for polychromatic thermal neutron radiation. Laboratory Directed Research and Development program of Idaho National Laboratory.

Defective Reduction in Frozen Pie Manufacturing Process

Science.gov (United States)

Nooted, Oranuch; Tangjitsitcharoen, Somkiat

2017-06-01

The frozen pie production has a lot of defects resulting in high production cost. Failure mode and effect analysis (FMEA) technique has been applied to improve the frozen pie process. Pareto chart is also used to determine the major defects of frozen pie. There are 3 main processes that cause the defects which are the 1st freezing to glazing process, the forming process, and the folding process. The Risk Priority Number (RPN) obtained from FMEA is analyzed to reduce the defects. If RPN of each cause exceeds 45, the process will be considered to be improved and selected for the corrective and preventive actions. The results showed that RPN values decreased after the correction. Therefore, the implementation of FMEA technique can help to improve the performance of frozen pie process and reduce the defects approximately 51.9%.

El pie del adulto mayor

OpenAIRE

Parodi García, José Francisco

2005-01-01

Acta Médica Peruana es una publicación del Colegio Médico del Perú El estudio del pie y sus patologías es un capítulo muy importante de la Geriatría. Las patologías del pie no escapan las características comunes de muchos síndromes geriátricos como son el ser asumidos como "normales para la edad", ser poco valorados, menospreciados, no tratados, con gran impacto sobre la salud en general y la calidad de vida.

Post irradiation examination of RAF/M steels after fast reactor irradiation up to 33 dpa and < 340 C (ARBOR1). RAFM steels. Metallurgical and mechanical characterisation. Final report for TW2-TTMS-001b, D9

Energy Technology Data Exchange (ETDEWEB)

Petersen, C. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). EURATOM, Inst. fuer Materialforschung, Programm Kernfusion

2010-07-01

In an energy generating fusion reactor structural materials will be exposed to very high dpa-levels of about 100 dpa. Due to this fact and because fast reactor irradiation facilities in Europe are not available anymore, a reactor irradiation at the State Scientific Center of the Russian Federation with its Research Institute of Atomic Reactors (SSC RIAR), Dimitrovgrad, had been performed in the fast reactor BOR 60 with an instrumented test rig. This test rig contained tensile, impact and Low Cycle Fatigue type specimens used at FZK since many years. Samples of actual Reduced Activation Ferritic/Martensitic (RAF/M) -steels (e.g. EUROFER 97) had been irradiated in this reactor at a lower temperature (< 340 C) up to a damage of 33 dpa. This irradiation campaign was called ARBOR 1. Starting in 2003 one half of these irradiated samples were post irradiation examined (PIE) by tensile testing, low cycle fatigue testing and impact testing under the ISTC Partner Contract 2781p in the hot cells of SSC RIAR. In the post irradiation instrumented impact tests a significant increase in the Ductile to Brittle Transition Temperature as an effect of irradiation has been detected. During tensile testing the strength values are increasing and the strain values reduced due to substantial irradiation hardening. The hardening rate is decreasing with increasing damage level, but it does not show saturation. The low cycle fatigue behaviour of all examined RAF/M - steels show at total strain amplitudes below 1 % an increase of number of cycles to failure, due to irradiation hardening. From these post irradiation experiments, like tensile, low cycle fatigue and impact tests, radiation induced design data, e.g. for verification of design codes, can be generated.

Tratamiento Actual del Pie Zambo

OpenAIRE

Salom Taverner, M.; Mínguez Rey, María Fe

2015-01-01

El pie zambo es una deformidad congénita en la que el pie presenta una deformidad en equino, varo de retropié, aductus del antepié y cavo. Su tratamiento ha variado con el tiempo. Clásicamente se trataba mediante distintas técnicas quirúrgicas que permitían corregir todas las deformidades pero que se acompañaban de frecuentes complicaciones a corto y largo plazo. Actualmente el tratamiento más extendido de esta deformidad es el tratamiento conservador mediante la técnica de Ponset...

Corrections to the 148Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

International Nuclear Information System (INIS)

Suyama, Kenya; Mochizuki, Hiroki

2006-01-01

The value of the burnup is one of the most important parameters of samples taken by post-irradiation examination (PIE). Generally, it is evaluated by the Neodymium-148 method. Precise evaluation of the burnup value requires: (1) an effective fission yield of 148 Nd; (2) neutron capture reactions of 147 Nd and 148 Nd; (3) a conversion factor from fissions per initial heavy metal to the burnup unit GWd/t. In this study, the burnup values of the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken by the Japan Atomic Energy Research Institute, were re-evaluated using more accurate corrections for each of these three items. The PIE data were then re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. The re-evaluation of the effective fission yield of 148 Nd has an effect of 1.5-2.0% on burnup values. Considering the neutron capture reactions of 147 Nd and 148 Nd removes dependence of C/E values of 148 Nd on the burnup value. The conversion factor from FIMA(%) to GWd/t changes according to the burnup value. Its effect on the burnup evaluation is small for samples having burnup of larger than 30 GWd/t. The analyses using the corrected burnup values showed that the calculated 148 Nd concentrations and the PIE data is approximately 1%, whereas this was 3-5% in prior analyses. This analysis indicates that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected by 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is an approximately 0.6% change in PIE samples having the burnup of larger than 30 GWd/t. Finally, comparison between calculation results using a single pin-cell model and an assembly model is carried out. Because the results agreed with each other within a few percent, we concluded that the single pin-cell model is suitable for the analysis of PIE samples and that the underestimation of plutonium isotopes, which occurred in the previous analyses, does not result from a geometry

Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs.

Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

International Nuclear Information System (INIS)

1995-09-01

At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs

Corrections to the {sup 148}Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya [Fuel Cycle Facility Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)]. E-mail: suyama.kenya@jaea.go.jp; Mochizuki, Hiroki [Japan Research Institute, Limited, 16 Ichiban-cho, Chiyoda-ku, Tokyo 102-0082 (Japan)

2006-03-15

The value of the burnup is one of the most important parameters of samples taken by post-irradiation examination (PIE). Generally, it is evaluated by the Neodymium-148 method. Precise evaluation of the burnup value requires: (1) an effective fission yield of {sup 148}Nd; (2) neutron capture reactions of {sup 147}Nd and {sup 148}Nd; (3) a conversion factor from fissions per initial heavy metal to the burnup unit GWd/t. In this study, the burnup values of the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken by the Japan Atomic Energy Research Institute, were re-evaluated using more accurate corrections for each of these three items. The PIE data were then re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. The re-evaluation of the effective fission yield of {sup 148}Nd has an effect of 1.5-2.0% on burnup values. Considering the neutron capture reactions of {sup 147}Nd and {sup 148}Nd removes dependence of C/E values of {sup 148}Nd on the burnup value. The conversion factor from FIMA(%) to GWd/t changes according to the burnup value. Its effect on the burnup evaluation is small for samples having burnup of larger than 30 GWd/t. The analyses using the corrected burnup values showed that the calculated {sup 148}Nd concentrations and the PIE data is approximately 1%, whereas this was 3-5% in prior analyses. This analysis indicates that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected by 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is an approximately 0.6% change in PIE samples having the burnup of larger than 30 GWd/t. Finally, comparison between calculation results using a single pin-cell model and an assembly model is carried out. Because the results agreed with each other within a few percent, we concluded that the single pin-cell model is suitable for the analysis of PIE samples and that the underestimation of plutonium isotopes, which occurred in the previous

Recent improvements in PIE-techniques at the IFE hot-laboratory

International Nuclear Information System (INIS)

Jenssen, Haakon Kristian

1999-01-01

The PIE-techniques at the IFE nuclear fuel section are continuously improved through upgrading of equipment and methods, e.g. image handling techniques and components utilised in data acquisition and editing techniques. To improve the quality or spatial resolution of neutron radiographs the normal technique was complemented with another method. The dysprosium foil/X-ray film technique is now supplemented with a track-etch recorder consisting of a cellulose nitrate film. For further examination the cellulose nitrate film can be digitised to allow electronic image treatment. Promising results were obtained with this technique on neutron radiographs, namely higher spatial resolution compared to the normal technique, high contrast and sharp neutron radiography images. The traditional one-axial profilometry of fuel rods were modified so that diameter/bow measurements are possible at several angular orientations during one acquisition sequence. This extension is very useful in several ways, for instance the built-in data symmetry of the method is used to check the correctness of the measurement results. Diameter and bow measurements give in addition detailed and complete information of canning irregularities and fuel rod profiles. PIE data are collected and compiled in an image file. Inspection and examination of the file information is possible through an ideal user-interface, i.e. Adobe photoshop software with navigator possibilities. An example incorporating PIE data acquired from a fuel rod exposed in a loss of coolant experiment is utilised for illustration of the user-interface and some of its possibilities (author) (ml)

Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

Energy Technology Data Exchange (ETDEWEB)

Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rader, Jordan D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examination facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.

Pied de Pulse

NARCIS (Netherlands)

2016-01-01

This project has two aims. The first is to study and implement a fractal-like structure of circles inspired by Apollonian circles, combined with a pied de poule (houndstooth). The second aim is to push the integration of electric actuators in garments, using the power of algorithmic design and

Recent status of development and irradiation performance for plate type fuel elements with reduced 235U enrichment at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.F.; Hassel, H.W.

1984-01-01

According to the present state of development full size test fuel elements with the maximum uranium densities of 2,2 g U/cm 3 meat for UAlsub(x), 3,2 g U/cm 3 meat for U 3 O 8 and 4,8 g U/cm 3 meat for U 3 Si 2 can be fabricated at NUKEM in production scale. Special chemical procedures for the uranium recovery were developed ensuring an economic fuel fabrication process. The post irradiation examinations (PIE) of 12 UAlsub(x) (U density 2,2 g U/cm 3 meat) and U 3 O 8 (up to 3,1 g U/cm 3 meat) test plates irradiated in the ORR, Oak Ridge research reactor, were terminated. All 12 test plates show unobjectionable irradiation behavior. Extensive irradiation tests on full size fuel elements were performed. All inserted elements show perfect irradiation behavior. The PIE of the first HFR Petten U 3 O 8 fuel elements are in progress. The full size ORR U 3 Si 2 fuel elements with so far highest uranium density of 4,76 g U/cm 3 meat achieved a burnup of 50 % loss of 235 U up to May 1983. One element was withdrawn from the reactor for PIE, the second will be irradiated to a burnup of 75 % loss of 235 U. The further development is concentrated on Usub(x)Sisub(y) fuel with highest uranium density. U 3 Si miniplates with up to 6,1 g U/cm 3 meat are supplied meeting the required specification, U 3 Si miniplates with 6,7 g U/cm 3 are in fabrication. (author)

Irradiation history and final postirradiation data for IFA-432

International Nuclear Information System (INIS)

Lanning, D.D.

1986-11-01

The irradiation history and final postirradiation examination (PIE) results are presented for the Halden fuel assembly IFA-432. In particular, the three instrumented rods (Rods 2, 3, and 5) that continued to the highest burnup (46 MWd/kgM) are emphasized. The long survival rate of the fuel instrumentation allowed a continuous record to be kept of linear heat generation rate and fuel temperatures throughout the operating history of the asembly. Fission gas release was estimated based on a combination of gas retention and ceramographic data. The PIE data show that the postirradiation condition of the rods is qualitatively consistent with the power/temperature histories experienced. Fission gas release estimates are: Rod 2 (380 μm as-fabricated gap), 20-40%; Rod 5 (230 μm as-fabricated gap), 20-40%; and Rod 3 (76 μm as-fabricated gap), 0-20%

Proceedings of the third JAERI-KAERI joint seminar on post irradiation examination technology

International Nuclear Information System (INIS)

1999-09-01

Between the Department of JMTR of the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Cycle Research Group of the Korea Atomic Energy Research Institute (KAERI), it has been periodically carried out the collaboration on technical information exchange by specialists and scientists, under the Arrangement of the Implementation of Cooperative Research Program in the Field of Peaceful Uses of Nuclear between JAERI and KAERI. And JAERI-KAERI joint seminar has been held every three years. The 1st and 2nd JAERI-KAERI Joint Seminars were held in November 1992 at JAERI and in September 1995 at KAERI, respectively. The 3rd JAERI-KAERI Joint Seminar was held on 25 and 26 March, 1999 at the Oarai Research Establishment of JAERI. In this seminar, total participants of 84 were joined from JAERI, KAERI, Hanyang University, Japan Nuclear Cycle Development Institute, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. Contributed presentations were in three sessions; Current status and future perspectives on PIE, PIE techniques and Evaluation of PIE data. Re-assembling technique for JOYO fuel, Nd-YAG laser welding technique, grain boundary analysis using FEG-TEM, lift time estimation of PWR Rod Cluster Control Assembly (RCCA) rodlet and failure analysis of Korea Nuclear Power Plant (KNP) fuel have been widely noticed as topic items on PIE. And some comments from PIE user, were pointed out that the nano-PIE technique, the flexibility to ad-hoc demands on testing space or utilization, and the international collaboration were very important for the next generation's PIE. The 34 of the present papers are indexed individually. (J.P.N.)

Proceedings of the third JAERI-KAERI joint seminar on post irradiation examination technology

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-09-01

Between the Department of JMTR of the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Cycle Research Group of the Korea Atomic Energy Research Institute (KAERI), it has been periodically carried out the collaboration on technical information exchange by specialists and scientists, under the Arrangement of the Implementation of Cooperative Research Program in the Field of Peaceful Uses of Nuclear between JAERI and KAERI. And JAERI-KAERI joint seminar has been held every three years. The 1st and 2nd JAERI-KAERI Joint Seminars were held in November 1992 at JAERI and in September 1995 at KAERI, respectively. The 3rd JAERI-KAERI Joint Seminar was held on 25 and 26 March, 1999 at the Oarai Research Establishment of JAERI. In this seminar, total participants of 84 were joined from JAERI, KAERI, Hanyang University, Japan Nuclear Cycle Development Institute, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. Contributed presentations were in three sessions; Current status and future perspectives on PIE, PIE techniques and Evaluation of PIE data. Re-assembling technique for JOYO fuel, Nd-YAG laser welding technique, grain boundary analysis using FEG-TEM, lift time estimation of PWR Rod Cluster Control Assembly (RCCA) rodlet and failure analysis of Korea Nuclear Power Plant (KNP) fuel have been widely noticed as topic items on PIE. And some comments from PIE user, were pointed out that the nano-PIE technique, the flexibility to ad-hoc demands on testing space or utilization, and the international collaboration were very important for the next generation's PIE. The 34 of the present papers are indexed individually. (J.P.N.)

Obesidad infantil: factor de riesgo para desarrollar pie plano

OpenAIRE

Saldívar-Cerón, Héctor Iván; Garmendia Ramírez, Alberto; Rocha Acevedo, Marco Antonio; Pérez-Rodríguez, Pedro

2015-01-01

Introducción: En México, la prevalencia de sobrepeso y obesidad en escolares es del 34.4% (ENSANUT 2012). Esta puede inducir un círculo vicioso “pie plano-plantalgia-sedentarismo-obesidad”. Sin embargo, la presencia y grado de pie plano en escolares con obesidad no se ha descrito en la población mexicana. El objetivo del estudio fue determinar la prevalencia de pie plano y su asociación con obesidad en escolares de Tamaulipas, México. Métodos: Se realizó un estudio analítico, transversal c...

Post-irradiation examinations and theoretical analyses of the fuel and fuel pin behavior of innovative systems; Bestrahlungsnachuntersuchungen und modelltheoretische Analysen zum Brennstoff- und Brennstabverhalten innovativer Systeme

Energy Technology Data Exchange (ETDEWEB)

Ernst, W.; Freund, D.; Geithoff, D.; Heck, M.; Jacobi, O.; Steiner, H.; Weimar, P.

1995-08-01

For the development of LMR cladding materials an irradiation test was performed in the KNK II reactor between 1986 and 1991. It demonstrated that a temperature gradient in the cladding has no observable effect on swelling behavior. PIEs of transient experiments in the Petten HFR have shown no significant cladding deformations. (orig.)

Improvements in PIE-techniques at the IFE hot-laboratory. 'Neutron radiography, three dimensional profilometry and image compilation of PIE data for visualization in an image based user-interface'

International Nuclear Information System (INIS)

Jenssen, H.K.; Oberlaender, B.C.

2002-01-01

The PIE-techniques used at IFE are continuously improved through upgrading of equipment and methods, e.g. image handling techniques and components utilized in data acquisition and editing techniques. To improve the quality or spatial resolution of neutron radiographs the normal technique was complemented with another method, i.e. the dysprosium foil/X ray film technique is supplemented with a track-etch recorder consisting of a cellulose nitrate film. For further examination of the neutron radiographs the cellulose nitrate film can be digitized to allow electronic image treatment. Promising results were obtained with this technique on neutron radiographs, namely higher spatial resolution compared to the normal technique, high contrast and sharp neutron radiography images. The traditional uniaxial profilometry of fuel rods was modified so that diameter/bow measurements are possible at several angular orientations during one acquisition sequence. This extension is very useful in several ways, for instance the built-in data symmetry of the method is used to check the correctness of the measurement results. Diameter and bow measurements give information of cladding irregularities and fuel rod profiles. Implementation of electronic image handling techniques is particularly useful in PIE when data are collected and compiled in an image file. Inspection and examination of the file contents (examination results) are possible through an ideal user-interface, i.e. Adobe Photoshop software with navigator possibilities. Examples incorporating PIE data acquired from neutron radiography, visual inspection and ceramography are utilized for illustration of the user-interface and some of its possibilities. (author)

Personalized Information Environments: Do Public Libraries Want a Slice of the PIE?

Science.gov (United States)

Worcester, Lea

2003-01-01

Discussion of personalized information environments (PIEs) focuses on library personalized interfaces. Describes academic library PIEs and commercial sites, such as Amazon.com that use personalization; explores the Public Library of Charlotte and Mecklenburg's personalized interface; and discusses opportunities and challenges PIEs offer public…

Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

Energy Technology Data Exchange (ETDEWEB)

Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

2013-12-15

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

International Nuclear Information System (INIS)

Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

2013-01-01

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

Solución Oportuna de Pie Plano

OpenAIRE

Espol; Tabares Cedeno, Maria Veronica; Ribadeneira Coloma, Betsy Evelyn; Zurita Aviles, Salomon Efrain

2015-01-01

Proponer una nueva técnica, donde se obtenga como beneficio la reparación definitiva de una lesion muy común como es el pie plano flexible y al mismo tiempo consolidar una empresa rentable y que genere valor ala sociedad el indice de prevalencia con diagnóstico de pie plano en el Ecuador es la especialidad de traumatología presentan este tipo de lesión. Guayaquil Magíster en Gerencia Hospitalaria

Pied de pulse: packing embroidered circles and coil actuators in pied de poule (houndstooth)

NARCIS (Netherlands)

Feijs, L.M.G.; Toeters, M.J.; Torrence, E.; Torrence, B.; Sequin, C.; McKenna, D.; Fenyvesi, K.; Sarhangi, R.

2016-01-01

We report on the integration of fractal-like structures inspired by Appolonian circles with pied de poule (houndstooth) fabric patterns, and on the use of electric actuators in garments made with digital embroidery techniques.

El pie en la danza clásica_

Directory of Open Access Journals (Sweden)

Nuria Massó Ortigosa.

2010-05-01

Full Text Available La danza es una forma de lenguaje que utiliza como instrumento al propio cuerpo, y su práctica continuada genera cambios y adaptaciones debidas a las altas exigencias de esta actividad. En este artículo se analizan las demandas, patologías y exigencias osteo-musculares del pie en la danza clásica. Para ello se ha partido del análisis de las estructuras del pie en distintas posiciones técnicas como la punta, media punta, o el “en dehors”. Como consecuencia de estas demandas de esfuerzo, se observa que una alta proporción de bailarines/as presenta alteraciones en el pie como hallux valgus y dedos en garra. También se ha estudiado la implicación en las técnicas básicas del pie de danza clásica de los músculos tríceps sural, peroneo lateral largo, tibial posterior, flexor largo y flexor corto del dedo gordo, tibial anterior, extensor propio del dedo gordo, abductor del dedo gordo, así como la musculatura interósea y plantar.

Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment

Energy Technology Data Exchange (ETDEWEB)

Collin, Blaise P.; Demkowicz, Paul A.; Baldwin, Charles A.; Harp, Jason M.; Hunn, John D.

2016-11-01

The PARFUME (PARticle FUel ModEl) code was used to predict silver release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO2) kernels. Post-irradiation examination (PIE) measurements were performed to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO2 compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude.

La ortopodología en el pie de riesgo (1ª Parte)

OpenAIRE

Dorca Coll, Adelina; Céspedes Céspedes, Tomás; Concustell Gonfaus, Josep; Cuevas Gómez, Rafael; Sacristán Valero, Sergi

1994-01-01

Los autores definen por primera vez el concepto de 'pie de riesgo' y ofrecen diferentes alternativas podológicas para prevenir lesiones en el pie y en el supuesto de que estas lesiones estén establecidas, presentan tratamientos ortésicos que equilibren el pie y eviten mayores cambios degenerativos.

Post-irradiation handling and examination at the HFEF complex

International Nuclear Information System (INIS)

Bacca, J.P.

1980-01-01

The Hot Fuel Examination Facility provides postirradiation handling and examination of fast reactor irradiation experiments and safety tests for the United States Breeder Reactor Program. Nondestructive interim examinations and destructive terminal examinations at HFEF derive data from tests irradiated in the Experimental Breeder Reactor No. II, in the Transient Reactor Test Facility (TREAT), and in the Sodium Loop Safety Facility. Similar support will be provided in the near future for tests irradiated in the Fast Flux Test Facility, and for the larger sodium loops to be irradiated in TREAT

RECH-1 test fuel irradiation status report

International Nuclear Information System (INIS)

Marin, J.; Lisboa, J.; Olivares, L.; Chavez, J.

2005-01-01

Since May 2003, one RECH-1 fuel element has been submitted to irradiation at the HFR-Petten, Holland. By November 2004 the irradiation has achieved its pursued goal of 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. This report presents the objectives and the current results of this fuel qualification under irradiation. Besides, a brief description of CHI/4/021, IAEA's Technical Cooperation Project that has supported this irradiation test, is also presented here. (author)

Technique of manufacturing specimen of irradiated fuel rods

International Nuclear Information System (INIS)

Min, Duck Seok; Seo, Hang Seok; Min, Duck Kee; Koo, Dae Seo; Lee, Eun Pyo; Yang, Song Yeol

1999-04-01

Technique of manufacturing specimen of irradiated fuel rods to perform efficient PIE is developed by analyzing the relation between requiring time of manufacturing specimen and manufacturing method in irradiated fuel rods. It takes within an hour to grind 1 mm of specimen thickness under 150 rpm in speed of grinding, 600 g gravity in force using no.120, no.240, no.320 of grinding paper. In case of no.400 of grinding paper, it takes more an hour to grind the same thickness as above. It takes up to a quarter to grind 80-130 μm in specimen thickness using no.400 of grinding paper. When grinding time goes beyond 15 minutes, the grinding thickness of specimen does not exist. The polishing of specimen with 150 Rpms in speed of grinding machine, 600 g gravity in force, 10 minutes in polishing time using diamond paste 15 μm on polishing cloths amounts to 50 μm in specimen thickness. In case of diamond paste 9 μm on polishing cloth, the polishing of specimen amounts to 20 μm. The polishing thickness of specimen with 15 minutes in polishing time using 6 μm, 3 μm, 1 μm, 1/4 μm does not exist. Technique of manufacturing specimen of irradiated fuel rods will have application to the destructive examination of PIE. (author). 6 refs., 1 tab., 10 figs

The post irradiation examination of fuel in support of Bruce A nuclear division fueling with flow program

International Nuclear Information System (INIS)

Montin, J.; Sagat, S.; Day, R.; Novak, J.; Bromfield, H.

1995-01-01

Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the assembly welds (endplateto-endcap welds) of all six cascaded bundles. No incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent endplate cracking. Also, an ultrasonic endplate inspection tool (UT) was developed and located in the fuel bay. to inspect fuelbundle endplates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unit 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assembly welds of fuel elements from Unit 4 (new outlet shield

The post-irradiation examination of fuel in support of Bruce A Nuclear Division fueling with flow program

International Nuclear Information System (INIS)

Montin, J.; Sagat, S.

1995-10-01

Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position of 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent end plate cracking. Also, an ultrasonic end plate inspection tool (UT) was developed and located in the fuel bay, to inspect fuel-bundle end plates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unite 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assemble welds of fuel elements from Unit 4 (new outlet shield-supported fuel bundles) confirming the UT results. (author). 5 refs., 8 figs

Constructing and applying the fractal pied de poule (houndstooth)

NARCIS (Netherlands)

Feijs, L.M.G.; Toeters, M.J.; Hart, G.; Sarhangi, R.

2013-01-01

Time is ready for a fractal version of pied de poule; it is almost "in the air". Taking inspiration from the Cantor set, and using the analysis of the classical pattern, we obtain a family of elegant new fractal Pied de Poules. We calculate the fractal dimension and develop an attractive fashion

16-rod-bundle: Irradiation in the MZFR and post-irradiation examinations

International Nuclear Information System (INIS)

Manzel, R.

1979-04-01

In the course of the irradiation of a 16-rod prototype bundle, the basis has been established for the irradiation of experimental fuel assemblies containing full-length PWR fuel rods in standard positions of the MZFR. The prototype bundle was discharged after an irradiation time of 284 full power days and a burnup of 11400 MWd/tU. The overall performance of the prototype bundle was highly satisfactory. Detailed post-irradiation examinations confirmed the good conditions of bundle structures and fuel rods. (orig.) [de

Segmented fuel irradiation program: investigation on advanced materials

International Nuclear Information System (INIS)

Uchida, H.; Goto, K.; Sabate, R.; Abeta, S.; Baba, T.; Matias, E. de; Alonso, J.

1999-01-01

The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

Safety Testing of AGR-2 UCO Compacts 6-4-2 and 2-3-1

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burns, Zachary M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

Post-irradiation examination (PIE) and elevated-temperature safety testing are being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). Details on this irradiation experiment have been previously reported [Collin 2014]. The AGR-2 PIE effort builds upon the understanding acquired throughout the AGR-1 PIE campaign [Demkowicz et al. 2015] and is establishing a database for the different AGR-2 fuel designs.

The construction of irradiated material examination facility

International Nuclear Information System (INIS)

Ro, Seung Gy; Lee, Key Soon; Herr, Young Hoi

1990-03-01

A detail design of the examination process, the hot cell facility and the annexed facility of the irradiated material examination facility (IMEF) which will be utilized to examine and evaluate physical and mechanical properties of neutron-irradiated materials, has been performed. Also a start-up work of the underground structure construction has been launched out. The project management and tasks required for the license application were duly carried out. The resultant detail design data will be used for the next step. (author)

Metallographic examination in irradiated materials examination facility

Energy Technology Data Exchange (ETDEWEB)

Choo, Yong Sun; Lee, Key Soon; Park, Dae Gyu; Ahn, Sang Bok; Yoo, Byoung Ok

1998-01-01

It is very important to have equipment of metallographic examination in hot-cell to observe the micro-structure of nuclear fuels and materials irradiated at nuclear power and/or research reactor. Those equipment should be operated by master-slave manipulators, so they are designed, manufactured and modified to make exercise easy and no trouble. The metallographic examination equipment and techniques as well as its operation procedure are described, so an operator can practice the metallography in hot-cell. (author). 5 refs., 7 tabs., 21 figs.

Proton irradiation effects on tensile and bend-fatigue properties of welded F82H specimens

Energy Technology Data Exchange (ETDEWEB)

Saito, S., E-mail: saito.shigeru@jaea.go.j [JAEA Tokai, J-PARC Center, 2-4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Kikuchi, K.; Hamaguchi, D. [JAEA Tokai, J-PARC Center, 2-4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Usami, K.; Ishikawa, A.; Nishino, Y.; Endo, S. [JAEA Tokai, Department of Hot Laboratories, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kawai, M. [KEK, Tsukuba-shi, Ibaraki-ken 305-0801 (Japan); Dai, Y. [PSI, Spallation Source Division, 5232 Villigen PSI (Switzerland)

2010-03-15

In several institutes, research and development for an accelerator-driven transmutation system (ADS) have been progressed. Ferritic/martensitic (FM) steels are the candidate materials for the beam window of ADS. To evaluate of the mechanical properties of the irradiated materials, the post irradiation examination (PIE) work of the SINQ (Swiss spallation neutron source) target irradiation program (STIP) specimens was carried out at JAEA. In present study, the results of PIE on FM steel F82H and its welded joint have been reported. The present irradiation conditions of the specimens were as follows: proton energy was 580 MeV. Irradiation temperatures were ranged from 130 to 380 deg. C, and displacement damage level was ranged from 5.7 to 11.8 dpa. The results of tensile tests performed at 22 deg. C indicated that the irradiation hardening occurred with increasing the displacement damage up to 10.1 dpa at 320 deg. C irradiation. At higher dose (11.8 dpa) and higher temperature (380 deg. C), irradiation hardening was observed, but degradation of ductility was relaxed in F82H welded joint. In present study, all specimens kept its ductility after irradiation and fractured in ductile manner. The results on bend-fatigue tests showed that the fatigue life (N{sub f}) of F82H base metal irradiated up to 6.3 dpa was almost the same with that of unirradiated specimens. The N{sub f} of the specimens irradiated up to 9.1 dpa was smaller than that of unirradiated specimens. Though the number of specimen was limited, the N{sub f} of F82H EB (15 mm) and EB (3.3 mm) welded joints seemed to increase after irradiation and the fracture surfaces of the specimens showed transgranular morphology. While F82H TIG welded specimens were not fractured by 10{sup 7} cycles.

The causal pie model: an epidemiological method applied to evolutionary biology and ecology.

Science.gov (United States)

Wensink, Maarten; Westendorp, Rudi G J; Baudisch, Annette

2014-05-01

A general concept for thinking about causality facilitates swift comprehension of results, and the vocabulary that belongs to the concept is instrumental in cross-disciplinary communication. The causal pie model has fulfilled this role in epidemiology and could be of similar value in evolutionary biology and ecology. In the causal pie model, outcomes result from sufficient causes. Each sufficient cause is made up of a "causal pie" of "component causes". Several different causal pies may exist for the same outcome. If and only if all component causes of a sufficient cause are present, that is, a causal pie is complete, does the outcome occur. The effect of a component cause hence depends on the presence of the other component causes that constitute some causal pie. Because all component causes are equally and fully causative for the outcome, the sum of causes for some outcome exceeds 100%. The causal pie model provides a way of thinking that maps into a number of recurrent themes in evolutionary biology and ecology: It charts when component causes have an effect and are subject to natural selection, and how component causes affect selection on other component causes; which partitions of outcomes with respect to causes are feasible and useful; and how to view the composition of a(n apparently homogeneous) population. The diversity of specific results that is directly understood from the causal pie model is a test for both the validity and the applicability of the model. The causal pie model provides a common language in which results across disciplines can be communicated and serves as a template along which future causal analyses can be made.

Assessments of sheath strain and fission gas release data from 20 years of power reactor fuel irradiations

International Nuclear Information System (INIS)

Purdy, P.L.; Manzer, A.M.; Hu, R.H.; Gibb, R.A.; Kohn, E.

1997-01-01

Over the past 20 years, many fuel elements or bundles discharged from Canadian CANDU power reactors have been examined in the AECL hot cells. The post-irradiation examination (PIE) database covers a wide range of operating conditions, from which fuel performance characteristics can be assessed. In the present analysis, a PIE database was compiled representing elements from a total of 129 fuel bundles, of which 26% (34 bundles) were confirmed to have one or more defective elements. This comprehensive database was assessed in terms of measured sheath strain and fission gas release (FGR) for intact elements, in an attempt to identify any changes in these parameters over the history of CANDU reactor operation. Results from this assessment indicate that, for the data that are typical of normal CANDU operating conditions, tensile sheath strain and FGR have remained within 0.5% and 8%, respectively. Those data beyond these ranges are from fuel operated under abnormal conditions, not representative of normal operation, and thus do not indicate a trend toward unexpected fuel behaviour. The distributions of the PIE measurements indicate that maximum expected sheath strains and FGR for normally operated fuel are 0.7% and 13%, respectively. (author)

HANARO fuel irradiation test(II)

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

Science.gov (United States)

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

Tritium release from beryllium pebbles after high temperature irradiation up to 3000 appm He in the HIDOBE-01 experiment

Energy Technology Data Exchange (ETDEWEB)

Til, S. van, E-mail: vantil@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Fedorov, A.V.; Stijkel, M.P.; Cobussen, H.L.; Mutnuru, R.K.; Idsert, P. van der [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and The Development of Fusion Energy, c/ Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

2013-11-15

In the HIDOBE (HIgh DOse irradiation of BEryllium) irradiation program, various grades of constrained and unconstrained beryllium pebbles, beryllium pellets and titanium-beryllide samples are irradiated in the High Flux Reactor (HFR) in Petten at four different temperatures (between 698 K and 1023 K) for 649 days [1]. The first of two HIDOBE irradiation experiments, HIDOBE-01, was completed after achieving a DEMO relevant helium production level of 3000 appm and the samples are retrieved for postirradiation examination (PIE). This work shows preliminary results of the out-of-pile tritium release analysis performed on different grades of irradiated beryllium pebbles (different in size). Relationships between irradiation temperature, tritium inventory and microstructural evolution have been observed by light microscopy and scanning electron microscopy.

Irradiation testing of coated particle fuel at Hanaro

International Nuclear Information System (INIS)

Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

2014-01-01

TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

New facility for post irradiation examination of neutron irradiated beryllium

International Nuclear Information System (INIS)

Ishitsuka, Etsuo; Kawamura, Hiroshi

1995-01-01

Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

Enfoque integral del paciente con pie diabético

Directory of Open Access Journals (Sweden)

Galia Fonseca Portilla

1996-04-01

Full Text Available Debe tenerse en cuenta que el diagnóstico de Diabetes Mellitus supone un cambio radical en las costumbres del paciente y que alcanzarlo puede requerir largo tiempo, por esto es necesario que los pacientes asistan a control cada dos o tres meses ye n cada visita insistirle en la importancia de curaciones. Si se continúa apoyando sobre la úlcera, se siguen exprimiendo, como esponja, las bacterias. Para el manejo del pie diabético se distinguen: prevención primaria (detección precoz del pie en riesgo, prevención secundaria (cuidado adecuado de las úlceras grado 1, 2 por el equipo interdisciplinario y prevención terciaria (a nivel hospitalario, para evitar la amputación de los pies con úlceras grado 3-5.

Structural Conservation and Functional Diversity of the Poxvirus Immune Evasion (PIE) Domain Superfamily.

Science.gov (United States)

Nelson, Christopher A; Epperson, Megan L; Singh, Sukrit; Elliott, Jabari I; Fremont, Daved H

2015-08-28

Poxviruses encode a broad array of proteins that serve to undermine host immune defenses. Structural analysis of four of these seemingly unrelated proteins revealed the recurrent use of a conserved beta-sandwich fold that has not been observed in any eukaryotic or prokaryotic protein. Herein we propose to call this unique structural scaffolding the PIE (Poxvirus Immune Evasion) domain. PIE domain containing proteins are abundant in chordopoxvirinae, with our analysis identifying 20 likely PIE subfamilies among 33 representative genomes spanning 7 genera. For example, cowpox strain Brighton Red appears to encode 10 different PIEs: vCCI, A41, C8, M2, T4 (CPVX203), and the SECRET proteins CrmB, CrmD, SCP-1, SCP-2, and SCP-3. Characterized PIE proteins all appear to be nonessential for virus replication, and all contain signal peptides for targeting to the secretory pathway. The PIE subfamilies differ primarily in the number, size, and location of structural embellishments to the beta-sandwich core that confer unique functional specificities. Reported ligands include chemokines, GM-CSF, IL-2, MHC class I, and glycosaminoglycans. We expect that the list of ligands and receptors engaged by the PIE domain will grow as we come to better understand how this versatile structural architecture can be tailored to manipulate host responses to infection.

Microstructural examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

2005-01-01

Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples from a Zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit 1, for a period for 6.77 effective full power years (EFPYs), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

Safety testing of AGR-2 UO₂ compacts 3-3-2 and 3-4-2

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

2015-09-01

Post-irradiation examination (PIE) is in progress on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2) [Collin 2014]. The AGR-2 PIE will build upon new information and understanding acquired throughout the recently-concluded six-year AGR-1 PIE campaign [Demkowicz et al. 2015] and establish a database for the different AGR-2 fuel designs.

Improving thermal model prediction through statistical analysis of irradiation and post-irradiation data from AGR experiments

International Nuclear Information System (INIS)

Pham, Binh T.; Hawkes, Grant L.; Einerson, Jeffrey J.

2014-01-01

As part of the High Temperature Reactors (HTR) R and D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test

Improving thermal model prediction through statistical analysis of irradiation and post-irradiation data from AGR experiments

Energy Technology Data Exchange (ETDEWEB)

Pham, Binh T., E-mail: Binh.Pham@inl.gov [Human Factor, Controls and Statistics Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hawkes, Grant L. [Thermal Science and Safety Analysis Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Einerson, Jeffrey J. [Human Factor, Controls and Statistics Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2014-05-01

As part of the High Temperature Reactors (HTR) R and D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test.

Management of spent fuel from research and prototype power reactors and residues from post-irradiation examination of fuel

International Nuclear Information System (INIS)

1989-09-01

The safe and economic management of spent fuel is important for all countries which have nuclear research or power reactors. It involves all aspects of the handling, transportation, storage, conditioning and reprocessing or final disposal of the spent fuel. In the case of spent fuel management from power reactors the shortage of available reprocessing capacity and the rising economic interest in the direct disposal of spent fuel have led to an increasing interest in the long term storage and management of spent fuel. The IAEA has played a major role in coordinating the national activities of the Member States in this area. It was against this background that the Technical Committee Meeting on ''Safe Management of Spent Fuel From Research Reactors, Prototype Power Reactors and Fuel From Commercial Power Reactors That Has Been Subjected to PIE (Post Irradiated Examination)'' (28th November - 1st December 1988) was organised. The aims of the current meeting have been to: 1. Review the state-of-the-art in the field of management of spent fuel from research and prototype power reactors, as well as the residues from post irradiation examination of commercial power reactor fuel. The emphasis was to be on the safe handling, conditioning, transportation, storage and/or disposal of the spent fuel during operation and final decommissioning of the reactors. Information was sought on design details, including shielding, criticality and radionuclide release prevention, heat removal, automation and remote control, planning and staff training; licensing and operational practices during each of the phases of spent fuel management. 2. Identify areas where additional research and development are needed. 3. Recommend areas for future international cooperation in this field. Refs, figs and tabs

Social suppliers: Exploring the cultural contours of the performance and image enhancing drug (PIED) market among bodybuilders in the Netherlands and Belgium.

Science.gov (United States)

van de Ven, Katinka; Mulrooney, Kyle J D

2017-02-01

This paper explores the understudied phenomenon of performance and image enhancing drug (PIED) markets by examining the structure and formation of the market for PIEDs among bodybuilders in the Netherlands and Belgium. This article specifically seeks to account for individual reasons and motivations for dealing PIEDs within these bodybuilding subcultures. Understanding illicit PIED markets is important for policy decisions as knowledge on the production and distribution of these substances may assist in designing law enforcement efforts, harm reduction initiatives and treatment options. This article draws on two years of fieldwork in various bodybuilding settings, 47 semi-structured interviews with individuals who are directly or indirectly involved in the PIED market and 64 PIED dealing cases initiated by criminal justice agencies in the Netherlands and Belgium. The data indicates that PIED dealing groups and individuals are often driven by motivations stemming from their social and cultural embeddedness in the bodybuilding subculture. Specifically, these PIED dealers are 'over-socialized' into the structure and culture of bodybuilding and follow the cultural scripts that come with their group affiliation and organization. As a result of the cultural context in which these transactions occur, PIED dealing networks among bodybuilders in the Netherlands and Belgium are more likely to consist of friends or 'friends of friends' tied together by threads of collective meaning found within the bodybuilding subculture. We argue that efforts seeking to explain the structure, formation and motivations of illicit PIED dealing must learn to appreciate how culture mediates structural forces and thereby influences individual and collective action. Policy makers, health care professionals and other relevant parties should consider a plurality of factors (social, economic and cultural) when designing and evaluating PIED-related interventions such as law enforcement efforts, harm

HANARO fuel irradiation test (II): revision

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

Annual report on operation, utilization and technical development of hot laboratories. From April 1, 1996 to March 31, 1997

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

This report describes activities, in fiscal year 1996, of the Reactor Fuel Examination Facility (RFEF), the Research Hot Laboratory (RHL) and the Waste Safety Testing Facility (WASTEF) which belong to the Department of Hot laboratories. In the RFEF, Post-Irradiation Examinations (PIEs) of PWR fuel assemblies irradiated in the Takahama Unit 3, a BWR fuel assembly irradiated in the Fukusima Daini Unit have been performed. Also, PIEs of assembly materials irradiated in the Fugen Reactor have been carried out. To support R and D works in JAERI, refabrication of segmented fuel rods have been done using irradiated LWR fuel rods for pulse irradiation in the NSRR and re-irradiation tests in the JMTR. PIEs have been performed on high burnup fuel rods and ROX fuel rods. For the RHL, PIEs have been performed on segment fuels irradiated in the NSRR, fuels and materials for HTTR, standard fuels for JRR-3M and materials for nuclear fusion reactor. In addition, a monitoring test of fuel elements in accordance with the surveillance program of the Magnox reactor of the Japan Atomic Power Corporation has been continued. In the WASTEF, leaching tests on TRU in simulated glass forms and a low flow rate tests on glass waste forms have been carried out. The examinations of alpha damage acceleration for the Synroc waste forms have also been performed. (author)

49 CFR 40.385 - Who bears the burden of proof in a PIE proceeding?

Science.gov (United States)

2010-10-01

... 49 Transportation 1 2010-10-01 2010-10-01 false Who bears the burden of proof in a PIE proceeding... TRANSPORTATION WORKPLACE DRUG AND ALCOHOL TESTING PROGRAMS Public Interest Exclusions § 40.385 Who bears the burden of proof in a PIE proceeding? (a) As the proponent of issuing a PIE, the initiating official bears...

Non-destructive tests of capsules for JMTR irradiation examination

International Nuclear Information System (INIS)

Tanaka, Hidetaka; Nagao, Yoshiharu; Sato, Masashi; Osawa, Kenji

2007-03-01

Irradiation examination are increasing in advanced irradiation research for accurate prediction control and evaluation of irradiation parameter such as neutron fluence, etc. by using JMTR. Irradiation capsule internals are therefore structurally complicated recently. This report described the procedure of non destructive tests such as radiographic test, penetrant test, ultrasonic test, etc. for inspection of irradiation capsules in JMTR, and the result of Test-case of confirmation procedure for internal parts of irradiation capsules. (author)

Shelf life of pie caps with biodegradable films as spacers

Directory of Open Access Journals (Sweden)

Daniela Verónica Escobar Gianni

2013-01-01

Full Text Available Commonly pie caps at market use polyethylene films as spacers between them. This paper studies the conventional spacers replacement with edible and biodegradable films made with whey protein isolate (WPI and potassium sorbate as a preservative. Besides facilitating the separation of pie caps, with this application is intended to increase their shelf life. The films made by the compression molding method were used as spacers in pie caps without preservative in their formula (A and with preservative (B and they were compared with conventional polyethylene spacers (C. During four months, monthly sensory, microbiological and physicochemical (humidity evaluations were done on the pie caps, together with humidity and solubility evaluations of the films. None of the samples showed microbiological or sensory deterioration. The sensory attributes showed no or slight difference in study time. Between samples the differences were minor: the best scores were for sample A in color, sample C in flavor, and samples B and C in texture and overall liking. The edible films have an interesting potential for this application, although studies in disguise the flavor of serum should be done.

Geophysical evidence for an extensive Pie de Palo Complex mafic-ultramafic belt, San Juan, Argentina

Science.gov (United States)

Chernicoff, Carlos J.; Vujovich, Graciela I.; van Staal, Cees R.

2009-12-01

The recent completion of a high-resolution aeromagnetic survey over the Pie de Palo uplift of the western Sierras Pampeanas has revealed an area of large magnetic anomalies associated with the Pie de Palo Complex. The Las Pirquitas thrust, which has transported and uplifted the Pie de Palo Complex, is recognized for at least 30 km in a roughly NE direction along the western boundary of the Pie de Palo Complex, beyond its limited outcrop. The type of sediments of the Caucete Group in the footwall of the Las Pirquitas thrust, which are regarded as the leading edge of the Precordillera terrane, are associated with much less pronounced magnetic anomalies. In addition, a conspicuous, NNE trending, broad magnetic high stands out in the survey, several kilometers to the east of the main outcrops of the Pie de Palo Complex; this broad magnetic anomaly bisects the Pie de Palo basement block, and continues further south at least as far as 32°S, the southern boundary of the latest aeromagnetic survey. This magnetic anomaly is interpreted to represent a structure corresponding to the Grenvillian Precordillera-Pie de Palo tectonic boundary zone, and would comprise the buried largest part of the mafic-ultramafic belt. The geophysical model of the magnetic data indicates that the boundary zone dips to the east, possibly suggesting the existence of a set of synthetic east dipping, west-verging thrusts, of which only one major structure (Las Pirquitas thrust) is exposed; the possibility of other slivers of upthrust boundary zone material cannot be excluded. It is considered that the Pie de Palo Complex represents a small sliver upthrust from the unexposed boundary zone material (containing highly magnetic mafic-ultramafic rocks). The east-dipping, west verging structures associated with the Pie de Palo Complex are suggested to represent an Ordovician reactivation of a Grenvillian suture zone developed when the Precordillera basement and Pie de Palo terrane docked; this

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

Energy Technology Data Exchange (ETDEWEB)

Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

2014-06-01

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

1999-01-01

The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

El pie del diabético

Directory of Open Access Journals (Sweden)

Parra Myriam

1990-12-01

Full Text Available

Existen complicaciones neurovasculares periféricas en el paciente diabético que de no prevenirse o tratarse precozmente pueden conducir a la formación de lesiones graves e irreversibles en el pie. Los grados de lesión pueden evolucionar desde la presencia de factores de riesgo hasta la formación de gangrena local o generalizada conllevando a la amputación del pie. La participación del paciente y la familia con relación a los cuidados podológicos, cumplimiento en el  tratamiento dietético, farmacológico, de ejercicio e identificación de riesgos son definitivos; estos aspectos requieren de la orientación e intervención del profesional de enfermería en el manejo de cada grado de lesión basándose en la identificación de diagnósticos reales y/o potenciales prioritarios.

Current status and future prospects of JMTR Hot Laboratory

International Nuclear Information System (INIS)

Baba, Osamu; Ooka, Norikazu; Hoshiya, Taiji

1999-01-01

A wide variety of post-irradiation examinations (PIEs) for research and development of nuclear fuels and materials to be utilized in nuclear field is available in three kinds of β - γ hot cells; concrete, lead and steel cells in the JMTR Hot Laboratory (JMTR HL) associated with the Japan Materials Testing Reactor (JMTR). In addition to PIEs, re-capsuling including re-instrumentation on the irradiated specimen is currently conducted for the power ramping tests of the LWR fuels using the Boiling Water Capsule (BOCA) or for the re-irradiation tests in the different neutron fields (coupling irradiation test). The newly developed techniques by the JMTR HL have provided us with the key information about the irradiation effects on mechanical and physical properties of the specimen in various environments as fission and fusion reactors. These techniques are focused on several topics as follows; (1) miniaturized specimen test as an advanced mechanical test, (2) slow strain rate tensile test (SSRT) and crack propagation measurement in high temperature and pressure water for the study of Irradiation Assisted Stress Corrosion Cracking (IASCC) of LWR core internals, (3) handling technique on materials containing tritium for the research and development of tritium breeders and neutron multiplier for fusion reactors, (4) jointing method using the conventional Tungsten Inert Gas (TIG) welding for re-assembling of irradiation capsules and/or re-fabrication of specimen, and (5) Nondestructive examination using ultrasonic wave and infrared thermography for the quantitative evaluation of irradiation embrittlement of structural materials in fission and fusion reactors. As there are various PIE facilities around Oarai site, mutual exchange of PIE information, interchange of researchers and mutual utilization on PIE facilities are desired to raise the scientific and technical potential on PIE and to get the break-through of the study in the field of nuclear applications. (author)

Irradiation Behavior and Post-Irradiation Examinations of an Acoustic Sensor Using a Piezoelectric Transducer

International Nuclear Information System (INIS)

Lambert, T.; Zacharie-Aubrun, I.; Hanifi, K.; Valot, Ch.; Fayette, L.; Rosenkantz, E.; Ferrandis, J.Y.; Tiratay, X.

2013-06-01

The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. In the framework of high burn-up fuel experiments under transient operating conditions, an innovative sensor based on acoustic method was developed by CEA and IES (Southern Electronic Institute).This sensor is used to determine the on-line composition of the gases located in fuel rodlet free volume and thus, allows calculating the molar fractions of fission gases and helium. The main principle of the composition determination by acoustic method consists in measuring the time of flight of an acoustic signal emitted and reflected in a specific cavity. A piezoelectric transducer, driven by a pulse generator, generates the acoustic wave in the cavity. The piezoelectric transducer is a PZT ceramic disk, mainly consisting of lead, zirconium and titanium. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. However, during the irradiation test, acoustic signal degradation was observed, mainly due to irradiation effect but also due to the increasing of the gas temperature. Despite this acoustic signal degradation, the time of flight measurements were carried out with good accuracy throughout the test, thanks to the development of a more efficient signal processing. After experiment, neutronic calculations were performed in order to determine neutron fluence at the level of the piezoelectric transducer. In order to have a better understanding of the acoustic sensor behavior under irradiation, Post Irradiation Examination program was done on piezoelectric transducer and on acoustic coupling material too. These examinations were also realized on a non-irradiated acoustic sensor built in the same conditions and with the same materials and the same

The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

International Nuclear Information System (INIS)

Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

1995-01-01

This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

Post-irradiation examinations of THERMHET composite fuels for transmutation

Science.gov (United States)

Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

2003-07-01

The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

Post-irradiation examinations of THERMHET composite fuels for transmutation

Energy Technology Data Exchange (ETDEWEB)

Noirot, J. E-mail: jnoirot@cea.fr; Desgranges, L.; Chauvin, N.; Georgenthum, V

2003-07-01

The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl{sub 2}O{sub 4} spinel inert matrix and around 40% weight of UO{sub 2} to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

Post-irradiation examinations of THERMHET composite fuels for transmutation

International Nuclear Information System (INIS)

Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

2003-01-01

The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2 O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour

La ortopodología en el pie de riesgo (2ª Parte)

OpenAIRE

Céspedes Céspedes, Tomás; Dorca Coll, Adelina; Concustell Gonfaus, Josep; Cuevas Gómez, Rafael; Sacristán Valero, Sergi

1994-01-01

Los autores presentan y describen varios casos clínicos podológicos, en los que aprecia: Una alteración y limitació progresiva de las funciones del pie y como resultado un transtomo general del equilibrio estático y dinámico: características que corresponden al ya descrito «Pie de riesgo'

El dolor patelofemoral en el baile flamenco y su relación con el pie

Directory of Open Access Journals (Sweden)

José M. Castillo-López

2017-05-01

Full Text Available El dolor patelofemoral es uno de los desórdenes más comunes que afectan a la rodilla. De las lesiones descritas en el baile flamenco profesional, destacan las relacionadas con el pie y la rodilla, en relación al microtraumatismo repetido del gesto técnico del zapateado y a la flexión de rodilla mantenida que requiere. El objetivo de este estudio es determinar la presencia de dolor patelo femoral en profesionales del baile de flamenco y comprobar si existe relación alguna con el pie pronador y genus valgo. En el estudio han participado 46 profesionales del baile flamenco, 38 bailaoras y 8 bailaores, con una carga media de 25.4 ± 8.25 horas de baile semanales. El 34.78% de los participantes presentaban dolor patelofemoral. El 24% presentaba también genus valgo. Se encontró un 34.78% de pies pronadores. La asociación de pies pronadores y genus valgo fue del 31.25%, mientras que la asociación de pies pronadores y dolor patelofemoral fue de 43.75 % de los casos. Factores predisponentes de esta patología se dan en la práctica del baile flamenco: el impacto repetitivo, una posición mantenida y forzada de la rodilla en un rango concreto, desplazamientos y los giros bruscos. Como conclusión, se demuestra relación directa entre la presencia de pies pronadores y dolor patelofemoral en el baile flamenco. Así como entre pies pronadores y genus valgo. Estudios científicos actuales demuestran la utilidad de soportes plantares en este síndrome. La incidencia de pies pronadores y dolor patelofemoral hace recomendable estudios biomecánicos y podológicos y el uso de soportes plantares personalizados en cada caso.

Postirradiation examination results for the Irradiation Effects Scoping Test 2

International Nuclear Information System (INIS)

Mehner, A.S.

1977-01-01

The postirradiation examination results are reported for two rods from the second scoping test (IE-ST-2) of the Nuclear Regulatory Commission Irradiation Effects Program. The rods were irradiated in the in-pile test loop of the Power Burst Facility at the Idaho National Engineering Laboratory. Rod IE-005 was fabricated from fresh fuel and cladding previously irradiated in the Saxton Reactor. Rod IE-006, fabricated from fresh fuel and unirradiated cladding, was equipped with six developmental cladding surface thermocouples. The rods were preconditioned, power ramped, and then subjected to film boiling operation. The performance of the rods and the developmental thermocouples are evaluated from the post irradiation examination results. The effects of prior irradiation damage in cladding are discussed in relation to fuel rod behavior during a power ramp and subsequent film boiling operation

PIE DIABÉTICO: ¿PODEMOS PREVENIRLO?

Directory of Open Access Journals (Sweden)

Dra. Oriana Paiva M.

2016-03-01

Full Text Available La diabetes es una enfermedad epidémica, con más de 400 millones de personas diagnosticadas en la actualidad. Las úlceras del pie son una de las complicaciones crónicas más frecuentes de esta población, con un riesgo acumulado de desarrollarla a lo largo de la vida de 15%. La neuropatía y la enfermedad vascular son las principales causas de su aparición. Las amputaciones no traumáticas en diabéticos son 15 veces más frecuentes que en la población general, superando en 2 a 3 veces la tasa en hombres que en mujeres. El 80% se puede prevenir. La clave para ello incluye un examen periódico del pie a todo diabético al menos 1 vez al año, educación del paciente en autocuidado, para que logre identificar precozmente la aparición de lesiones evitando mayores complicaciones, un zapato adecuado y un buen control metabólico.

LOCA testing of high burnup PWR fuel in the HBWR. Additional PIE on the cladding of the segment 650-5

Energy Technology Data Exchange (ETDEWEB)

Oberlaender, B.C.; Espeland, M.; Jenssen, H.K.

2008-07-01

IFA-650.5, a test with pre-irradiated fuel in the Halden Project LOCA test series, was conducted on October 23rd, 2006. The fuel rod had been used in a commercial PWR and had a high burnup, 83 MWd/kgU. Experimental arrangements of the fifth test were similar to the preceding LOCA tests. The peak cladding temperature (PCT) level was higher than in the third and fourth tests, 1050 C. A peak temperature close to the target was achieved and cladding burst occurred at approx. 750 C. Within the joint programme framework of the Halden Project PIE was done, consisting of gamma scanning, visual inspection, neutron-radiography, hydrogen analysis and metallography / ceramography. An additional extensive PIE including metallography, hydrogen analysis, and hardness measurements of cross-sections at seven axial elevations was done. It was completed to study the high burnup and LOCA induced effects on the Zr-4 cladding, namely the migration of oxygen into the cladding from the inside surface, the cladding distension, and the burst (author)(tk)

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

Development of a Scanning Microscale Fast Neutron Irradiation Platform for Examining the Correlation Between Local Neutron Damage and Graphite Microstructure

Energy Technology Data Exchange (ETDEWEB)

Pinhero, Patrick [Univ. of Missouri, Columbia, MO (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-10

The fast particle radiation damage effect of graphite, a main material in current and future nuclear reactors, has significant influence on the utilization of this material in fission and fusion plants. Atoms on graphite crystals can be easily replaced or dislocated by fast protons and result in interstitials and vacancies. The currently accepted model indicates that after most of the interstitials recombine with vacancies, surviving interstitials form clusters and furthermore gather to create loops with each other between layers. Meanwhile, surviving vacancies and interstitials form dislocation loops on the layers. The growth of these inserted layers cause the dimensional increase, i.e. swelling, of graphite. Interstitial and vacancy dislocation loops have been reported and they can easily been observed by electron microscope. However, observation of the intermediate atom clusters becomes is paramount in helping prove this model. We utilize fast protons generated from the University of Missouri Research Reactor (MURR) cyclotron to irradiate highly- oriented pyrolytic graphite (HOPG) as target for this research. Post-irradiation examination (PIE) of dosed targets with high-resolution transmission electron microscopy (HRTEM) has permit observation and analysis of clusters and dislocation loops to support the proposed theory. Another part of the research is to validate M.I. Heggie’s Ruck and Tuck model, which introduced graphite layers may fold under fast particle irradiation. Again, we employed microscopy to image irradiated specimens to determine how the extent of Ruck and Tuck by calculating the number of folds as a function of dose. Our most significant accomplishment is the invention of a novel class of high-intensity pure beta-emitters for long-term lightweight batteries. We have filed four invention disclosure records based on the research conducted in this project. These batteries are lightweight because they consist of carbon and tritium and can be

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

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

Post-irradiation examinations of inert matrix nitride fuel irradiated in JMTR (01F-51A capsule)

International Nuclear Information System (INIS)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Honda, Junichi; Hatakeyama, Yuichi; Ono, Katsuto; Matsui, Hiroki; Arai, Yasuo

2007-03-01

A plutonium nitride fuel pin containing inert matrix such as ZrN and TiN was encapsulated in 01F-51A and irradiated in JMTR. Minor actinides are surrogated by plutonium. Average linear powers and burnups were 408W/cm, 30000MWd/t(Zr+Pu) [132000MWd/t-Pu] for (Zr,Pu)N and 355W/cm, 38000MWd/t(Ti+Pu) [153000MWd/t-Pu] for (TiN,PuN). The irradiated capsule was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pin. Very low fission gas release rate of about 1.6% was measured. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

La enfermedad de manos, pies y boca (Hand, Foot, and Mouth Disease)

Centers for Disease Control (CDC) Podcasts

La enfermedad de manos, pies y boca es una enfermedad contagiosa que afecta principalmente a niños menores de cinco años. En este podcast, la doctora Eileen Schneider habla de los síntomas de la enfermedad de manos, pies y boca, cómo se propaga y cómo protegerse y proteger a sus hijos para que no se infecten con el virus.

21 CFR 152.126 - Frozen cherry pie.

Science.gov (United States)

2010-04-01

... incorporating in a filling contained in a pastry shell mature, pitted, stemmed cherries that are fresh, frozen, and/or canned. The top of the pie may be open or it may be wholly or partly covered with pastry or other suitable topping. Filling, pastry, and topping components of the food consist of optional...

Postirradiation examination of light water reactor fuel: a United States perspective

International Nuclear Information System (INIS)

Neimark, L.A.; Ocken, H.

1980-01-01

Poolside and hot-cell postirradiation examination (PIE) have played and will continue to play a significant role in the US LWR program. The principal uses of PIE are in fuel surveillance, fuel improvement, and failure analysis programs and in the postmortem analysis of safety-related tests. Institutional problems associated with fuel shipping, waste disposal, and fuel disposal can be expected to pose obstacles to hot-cell examinations and likely result in more sophisticated poolside examinations

PIE the search: searching PubMed literature for protein interaction information.

Science.gov (United States)

Kim, Sun; Kwon, Dongseop; Shin, Soo-Yong; Wilbur, W John

2012-02-15

Finding protein-protein interaction (PPI) information from literature is challenging but an important issue. However, keyword search in PubMed(®) is often time consuming because it requires a series of actions that refine keywords and browse search results until it reaches a goal. Due to the rapid growth of biomedical literature, it has become more difficult for biologists and curators to locate PPI information quickly. Therefore, a tool for prioritizing PPI informative articles can be a useful assistant for finding this PPI-relevant information. PIE (Protein Interaction information Extraction) the search is a web service implementing a competition-winning approach utilizing word and syntactic analyses by machine learning techniques. For easy user access, PIE the search provides a PubMed-like search environment, but the output is the list of articles prioritized by PPI confidence scores. By obtaining PPI-related articles at high rank, researchers can more easily find the up-to-date PPI information, which cannot be found in manually curated PPI databases. http://www.ncbi.nlm.nih.gov/IRET/PIE/.

Consideraciones anatómicas y funcionales en la patología del quinto radio del pie

OpenAIRE

Gallart Ortega, José; Lahoz Gimeno, Manuel

2013-01-01

OBJETIVO: Se analiza si las diferentes alteraciones morfológicas y funcionales del pie se asocian estadísticamente a las diferentes desviaciones del quinto radio en el plano transverso y/o sagital como posible causa o agravantes de ésta. A su vez se describen las diferentes patologías quirúrgicas en quinto radio del pie y se evalúa el resultado goniométrico postquirúrgico de las diferentes técnicas quirúrgicas empleadas. MATERIAL MÉTODOS: En una muestra consecutiva de 121 pies intervenidos qu...

Valoración de los resultados del tratamiento quirúrgico del pie plano estático

OpenAIRE

Paz Jiménez, José; Amigo Fernandez, A.; Romo Contreras, I.

1989-01-01

La deformidad fundamental del pie plano es el valgo de talón. Existe siempre un movimiento helicoidal en pronación del retropic y supinación del antepic. En este trabajo se hace una revisión de los pacientes tratados en nuestro Servicio a lo largo de diez años, que han sido sometidos a intervención quirúrgica, mayoritariamente artrodesis. - El fotopodograma constituye un excelente método para controlar la evolución de un pie plano. El pie plano por astrágalo vertical es irreductible des...

Results of postirradiation examination of the in-pile blockage experiments MOL-7C/4 and MOL-7C/5

International Nuclear Information System (INIS)

Weimar, P.; Schleisiek, K.

1991-01-01

The Mol-7C in-pile local blockage experiments are performed in the BR-2 reactor at Mol, Belgium as a joint project of Kernforchungszentrum Karlsruhe (KfK) and Studiecentrum voor Kernenergie/Centre d'Etude de l'Energie Nuclearire-Mol. The main objective is to investigate the consequences of local cooling disturbances in liquid-metal-cooled reactor (LMR) fuel subassemblies. In the tests Mol-7C/4 and MOL-7C/5, fuel pins from KNK II are used with a burnup of 5 and 1.7%, respectively. An active central porous blockage is used to simulate the cooling disturbance. During irradiation, the blockage causes significant local damage, including melting of cladding and fuel. Extensive postirradiation examinations (PIE) are performed to investigate the extent of damage. In this paper a description and interpretation of results of the destructive PIE performed at the Hot Cells Laboratory at KfK is given, along with some conclusions related to LMR safety

Irradiated fuel performance evaluation technology development

International Nuclear Information System (INIS)

Koo, Yang Hyun; Bang, J. G.; Kim, D. H.

2012-01-01

Alpha version performance code for dual-cooled annular fuel under steady state operation, so called 'DUOS', has been developed applying performance models and proposed methodology. Furthermore, nonlinear finite element module which could be integrated into transient/accident fuel performance code was also developed and evaluated using commercial FE code. The first/second irradiation and PIE test of annular pellet for dual-cooled annular fuel in the world have been completed. In-pile irradiation test DB of annular pellet up to burnup of 10,000 MWd/MTU through the 1st test was established and cracking behavior of annular pellet and swelling rate at low temperature were studied. To do irradiation test of dual-cooled annular fuel under PWR's simulating steady-state conditions, irradiation test rig/rod design/manufacture of mock-up/performance test have been completed through international collaboration program with Halden reactor project. The irradiation test of large grain pellets has been continued from 2002 to 2011 and completed successfully. Burnup of 70,000 MWd/MTU which is the highest burnup among irradiation test pellets in domestic was achieved

Geometry and computation of Houndstooth (Pied-de-poule)

NARCIS (Netherlands)

Feijs, L.M.G.; Bosch, Robert; McKenna, Douglas; Sarhangi, Reza

2012-01-01

We apply a variety of geometric and computational tools to improve our understanding of the Houndstooth (Pied de poule) pattern. Although the pattern must have been known for centuries, it was made famous mostly by Christian Dior and is still frequently used in many variations. It is a

Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

1984-10-01

Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

Factores asociados al pie diabético en pacientes ambulatorios. Centro de Diabetes Cardiovascular del Caribe. Barranquilla (Colombia

Directory of Open Access Journals (Sweden)

Mildred Rosales Amarís

2012-01-01

respectivamente, la neuropatía (OR = 10.14; p = 0.00 y la deformidad de los pies (OR = 10.02; p = 0.00 se asociaron significativamente al pie diabético. Se encontró como factor protector el autocuidado (OR = 0.21; p = 0.00000 y la educación relacionada con el control de factores de riesgo para la diabetes (OR = 0.32; p = 0.00007. Conclusión: La macroangiopatía, la microangiopatía, la neuropatía y la deformidad en los pies se comportaron como factores asociados a la presencia del pie diabético. El autocuidado contribuyó a la prevención del mismo.

Enlarging the Societal Pie Through Wise Legislation: A Psychological Perspective.

Science.gov (United States)

Baron, Jonathan; Bazerman, Max H; Shonk, Katherine

2006-06-01

We offer a psychological perspective to explain the failure of governments to create near-Pareto improvements. Our tools for analyzing these failures reflect the difficulties people have trading small losses for large gains: the fixed-pie approach to negotiations, the omission bias and status quo bias, parochialism and dysfunctional competition, and the neglect of secondary effects. We examine the role of human judgment in the failure to find wise trade-offs by discussing diverse applications of citizen and government decision making, including AIDS treatment, organ-donation systems, endangered-species protection, subsidies, and free trade. Our overall goal is to offer a psychological approach for understanding suboptimality in government decision making. © 2006 Association for Psychological Science.

Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

Energy Technology Data Exchange (ETDEWEB)

Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organization (Australia)

2009-06-15

Introduction: The RERTR program is responsible for converting research reactors that use high-enriched uranium fuels to ones that use low-enriched uranium fuels [1]. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. Based on the results of initial fuel plate testing, adjustments have been made to the characteristics of fuel plates to improve the stability of the fuel microstructure. One improvement has been to add Si to the matrix of a dispersion fuel. This material is also being added at the fuel/cladding interface of a monolithic fuel. This paper will discuss the irradiation performance of these fuels, in terms of the stability of their microstructures during irradiation. Results and discussion: The post-irradiation examinations of fuel plates are performed at the Idaho National Laboratory. These examinations consist of visual examinations of fuel plates, gamma scanning, thickness measurements, oxide thickness measurements, and optical metallographic examinations of the fuel plate microstructures. Microstructural analysis is also performed using scanning electron microscopy. Overall, U-7Mo and U-10Mo alloy fuels have displayed the best irradiation performance, particularly, when a Si-containing Al alloy is used as the dispersion fuel matrix. The benefit of using this type of matrix is that the commonly observed fuel/cladding interaction that occurs during irradiation is reduced and the interaction layer that forms exhibit stable behavior during irradiation. Monolithic-type fuels, which consist of a U-Mo foil encased in Al alloy cladding, are also being developed. These types of fuels are also showing promise and will continue to be developed. One challenge with this type of fuel is in trying to maximize the bond strength at the foil/cladding interface. Fuel/cladding interactions can affect the quality of the boding at this interface. Si is being added to improve the characteristics

Efectividad de un programa educativo en pacientes con pie diabético de riesgo

Directory of Open Access Journals (Sweden)

Olga Lidia Pereira Despaigne

2015-01-01

Full Text Available Se realizó un estudio de intervención terapéutica educativa en pacientes con pie diabético de riesgo, atendidos en la Consulta de Angiopatía Diabética del Hospital Provincial Docente Clinicoquirúrgico "Saturnino Lora Torres" de Santiago de Cuba, desde enero del 2011 hasta igual mes del 2013, a fin de determinar la efectividad de un programa educativo (momento de la intervención al respecto. Fueron indicativos de pie diabético de riesgo (momento del diagnóstico, la úlcera previa o amputación, la neuropatía periférica con pérdida de sensación protectora, la enfermedad arterial periférica, la hiperqueratosis y la afección grave de las uñas, la deformidad en el pie, la movilidad articular reducida, el calzado inadecuado, el tabaquismo, la edad avanzada, la larga evolución de la diabetes mellitus y la falta de educación diabetológica en pie diabético. La información fue analizada y procesada mediante las pruebas de hipótesis de Mann-Whitney y de comparación de medias, con un error de 5 %, y se obtuvo que con este programa (momento evaluativo se disminuyera la afectación que el pie diabético ocasiona a la actividad social, laboral y familiar de estas personas, así como el elevado costo económico, tanto para ellos como para el sistema de salud

Great tit and pied flycatcher populations on the territory of radioactive trail in the eastern Urals

International Nuclear Information System (INIS)

Lebedeva, N.V.

1994-01-01

The purpose of this preliminary study was to evaluate the prospects for using populations of the great tit (Parus major) and the pied flycatcher (Ficedula hypoleuca) in ecological monitoring of territories contaminated with radionuclides. Studies were performed in the summer of 1992 in birch forests of the Southern Urals. Artificial nesting sites (log houses) located 1.5 m above the ground were distributed over territories of two test plots: 30 log houses on the contaminated plot and 60 on the control plot (contamination with Sr-90 1.5 and 2 · 10 -3 mCi/m 2 , respectively). The nesting success in the great tit was similar on both plots, whereas that in the pied flycatcher was significantly greater on the control plot (in a open-quotes cleanclose quotes forest). Pied flycatchers build their nests out of highly radioactive materials, whereas great tits use nonradioactive or weakly emitting materials. Hence, pied flycatcher's nestlings receive a significant radiation dose from components of the nest: From the moment of egg laying, this dose amounts to 0.5 rem. The pied flycatcher can be used as an indicator of radioactive contamination. This species is more sensitive than the great tit, which has been successfully used as an indicator for monitoring the industrial air pollution

Channel Bow in Boiling Water Reactors - Hot Cell Examination Results and Correlation to Measured Bow

International Nuclear Information System (INIS)

Mahmood, S.T.; Lin, Y.P.; Dubecky, M.A.; Edsinger, K.; Mader, E.V.

2007-01-01

An increase in frequency of fuel channel-control blade interference has been observed in Boiling Water Reactors (BWR) in recent years. Many of the channels leading to interference were found to bow towards the control blade in a manner that was inconsistent with the expected bow due to other effects. The pattern of bow appeared to indicate a new channel bow mechanism that differed from the predominant bow mechanism caused by differential growth due to fast-fluence gradients. In order to investigate this new type of channel bow, coupons from several channels with varying degrees of bow were returned to the GE Vallecitos Nuclear Center (VNC) for Post-Irradiation Examination (PIE). This paper describes the characteristics of channel corrosion and hydrogen pickup observed, and relates the observations to the channel exposure level, control history, and measured channel bow. The channels selected for PIE had exposures in the range of 36-48 GWd/MTU and covered a wide range of measured bow. The coupons were obtained at 4 elevations from opposing channel sides adjacent and away from the control blade. The PIE performed on these coupons included visual examination, metallography, and hydrogen concentration measurements. A new mechanism of control-blade shadow corrosion-induced channel bow was found to correlate with differences in the extent of corrosion and corresponding differences in the hydrogen concentration between opposite sides of the channels. The increased corrosion on the control blade sides was found to be dependent on the level of control early in the life of the channel. The contributions of other potential factors leading to increased channel bow and channel-control blade interference are also discussed in this paper. (authors)

Caracterización clínico epidemiológica de pacientes afectados por pie diabético

Directory of Open Access Journals (Sweden)

Dania Núñez Álvarez

2018-01-01

Full Text Available Introducción: el pie diabético constituye un problema de salud en los pacientes portadores de diabetes mellitus y se registra como la primera causa de amputación no traumática de los miembros. El conocimiento de sus particularidades clínico-epidemiológicas resulta de sumo interés. Objetivo: caracterizar clínica y epidemiológicamente a los pacientes con pie diabético neuroinfeccioso en el Hospital "Dr. Joaquín Castillo Duany" de Santiago de Cuba. Método: se realizó un estudio descriptivo retrospectivo. El universo estuvo constituido por 96 pacientes diagnosticados con pie diabético durante el periodo comprendido entre el 2011 y el 2015. Se emplearon variables demográficas y relacionadas con factores de riesgo generales y específicos del pie diabético. Resultados: el pie diabético predominó entre la cuarta y sexta décadas de vida (24,7 % y en pacientes del sexo femenino (59,7 %. La diabetes mellitus tipo 2 estuvo representada en el 52,5 %. Mientras el factor de riesgo general de pie diabético más frecuente fue la obesidad (56,7 %, la neuropatía diabética, como factor de riesgo específico se presentó en el 98,9 % de los enfermos estudiados. Conclusiones: el pie diabético predominó en la cuarta y sexta décadas de la vida y en el sexo femenino. El mayor porcentaje de los pacientes era portador de diabetes mellitus tipo 2. La obesidad, la neuropatía diabética periférica, el sobrepeso, y la hiperglucemia constituyeron los factores de riesgo prevalecientes.

Post irradiation examination of garter springs from Indian PHWRs

International Nuclear Information System (INIS)

Dubey, J.S.; Shah, Priti Kotak; Mishra, Prerna; Singh, H.N.; Alur, V.D.; Kumar, Ashwini; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

2013-12-01

Irradiated Zr-2.5Nb-0.5Cu garter springs, belonging to Indian Pressurised Heavy Water Reactors, which had experienced 8 to 10 Effective Full Power Years of operation were subjected to visual, dimensional, chemical, metallographic examination and relevant mechanical tests. Methodology of the tests conducted and results are presented. The digital photographs were used to measure the inner and outer circumferences by image processing. The hydrogen (H) content in the spring coils were measured using Differential Scanning Calorimetry (DSC). In the stretch test, all the irradiated GSs were found to require an additional load, as compared to unirradiated GS, to produce a given amount of residual extension which indicated that the irradiated GSs had undergone significant irradiation hardening. The crush test results showed that the minimum load required to crush the coil or cause a sudden sideways shift in the grips was higher than 400 N/coil, much higher than the design load. The test results indicated that the irradiated GS, after 10 EFPY of operation, have adequate strength and ductility to continue to meet the design intent. Mechanical tests were carried out on irradiated girdle wires taken out of the loose fit garter springs (GS) from (NAPS-1, ∼ 8.5 EFPY) and tight fit garter spring from KAPS-2 (∼ 8.0 EFPY) PHWRs. Tensile tests on the irradiated girdle wires, showed irradiation hardening in the material and reduction in ductility. The irradiated girdle wires have around 4 to 5% residual ductility level against the 15% ductility of unirradiated wire. The fracture surfaces of the irradiated as well as the un-irradiated girdle wires were observed in SEM. (author)

A re-examination of thermodynamic modelling of U-Ru binary phase diagram

Energy Technology Data Exchange (ETDEWEB)

Wang, L.C.; Kaye, M.H., E-mail: matthew.kaye@uoit.ca [University of Ontario Institute of Technology, Oshawa, ON (Canada)

2015-07-01

Ruthenium (Ru) is one of the more abundant fission products (FPs) both in fast breeder reactors and thermal reactors. Post irradiation examinations (PIE) show that both 'the white metallic phase' (MoTc-Ru-Rh-Pd) and 'the other metallic phase' (U(Pd-Rh-Ru)3) are present in spent nuclear fuels. To describe this quaternary system, binary subsystems of uranium (U) with Pd, Rh, and Ru are necessary. Presently, only the U-Ru system has been thermodynamically described but with some problems. As part of research on U-Ru-Rh-Pd quaternary system, an improved consistent thermodynamic model describing the U-Ru binary phase diagram has been obtained. (author)

Post-irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (89F-3A capsule)

International Nuclear Information System (INIS)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Arai, Yasuo; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki

2000-03-01

Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 89F-3A and irradiated in JMTR up to 5.5% FIMA at a maximum linear power of 73 kW/m. The capsule cooled for ∼5 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. Very low fission gas release rate of about 2 ∼ 3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

TEM examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

2005-09-01

In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

The pied piper of Hamelin: Exploring the story

Directory of Open Access Journals (Sweden)

Leila Mury Bergmann

2008-12-01

Full Text Available This article’s goal is to explore the story of The Pied Piper of Hamelin. In order to do so, social aspects of childhood in the (probable time when the story was created are presented, as well as some curiosities about the text. Furthermore, this paper brings a personal statement (by the authors of a storytelling activity in a public school nursery in Porto Alegre, based upon the version of the book The Pied Piper of Hamelin, by Tatiana Belinky. After some questions and informal conversation about the tale, the children’s verbal expositions were analyzed. The research brings, as a theoretical background, contributions of studies from authors connected, mainly, to the field of Children’s Literature and Education. It brought the conclusion that the piper’s story motivated representative interventions, since during the storytelling activity children used typical aspects of their culture and also showed their dominium of temporal arrangement of the text’s events, as well as their peculiar creativity – according to their age – of recreating in the situation of the so-called make-believe.

9 CFR 381.158 - Poultry dinners (frozen) and pies.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Poultry dinners (frozen) and pies. 381... AGRICULTURE AGENCY ORGANIZATION AND TERMINOLOGY; MANDATORY MEAT AND POULTRY PRODUCTS INSPECTION AND VOLUNTARY INSPECTION AND CERTIFICATION POULTRY PRODUCTS INSPECTION REGULATIONS Definitions and Standards of Identity or...

Survey of post-irradiation examinations made of mixed carbide fuels

International Nuclear Information System (INIS)

Coquerelle, M.

1997-01-01

Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

Microstructural examination of irradiated vanadium alloys

Energy Technology Data Exchange (ETDEWEB)

Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States); Chung, H.M. [Argonne National Lab., IL (United States)

1997-04-01

Microstructural examination results are reported for a V-5Cr-5Ti unirradiated control specimens of heat BL-63 following annealing at 1050{degrees}C, and V-4Cr-4Ti heat BL-47 irradiated in three conditions from the DHCE experiment: at 425{degrees}C to 31 dpa and 0.39 appm He/dpa, at 600{degrees}C to 18 dpa and 0.54 appm He/dpa and at 600{degrees}C to 18 dpa and 4.17 appm He/dpa.

prepare_taxa_charts.py: A Python program to automate generation of publication ready taxonomic pie chart images from QIIME.

Science.gov (United States)

Lakhujani, Vijay; Badapanda, Chandan

2017-06-01

QIIME (Quantitative Insights Into Microbial Ecology) is one of the most popular open-source bioinformatics suite for performing metagenome, 16S rRNA amplicon and Internal Transcribed Spacer (ITS) data analysis. Although, it is very comprehensive and powerful tool, it lacks a method to provide publication ready taxonomic pie charts. The script plot_taxa_summary . py bundled with QIIME generate a html file and a folder containing taxonomic pie chart and legend as separate images. The images have randomly generated alphanumeric names. Therefore, it is difficult to associate the pie chart with the legend and the corresponding sample identifier. Even if the option to have the legend within the html file is selected while executing plot_taxa_summary . py , it is very tedious to crop a complete image (having both the pie chart and the legend) due to unequal image sizes. It requires a lot of time to manually prepare the pie charts for multiple samples for publication purpose. Moreover, there are chances of error while identifying the pie chart and legend pair due to random alphanumeric names of the images. To bypass all these bottlenecks and make this process efficient, we have developed a python based program, prepare_taxa_charts . py , to automate the renaming, cropping and merging of taxonomic pie chart and corresponding legend image into a single, good quality publication ready image. This program not only augments the functionality of plot_taxa_summary . py but is also very fast in terms of CPU time and user friendly.

Preliminary test results for post irradiation examination on the HTTR fuel

International Nuclear Information System (INIS)

Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

2007-01-01

The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

In-pile IASCC growth tests of irradiated stainless steels in JMTR

Energy Technology Data Exchange (ETDEWEB)

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Shibata, Akira; Ohmi, Masao [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation-assisted stress corrosion cracking (IASCC) test plan to evaluate in-situ effects of neutron/{gamma}-ray irradiation on crack growth of irradiated stainless steels under high-temperature water conditions for commercial boiling water reactors (BWRs) using the Japan Materials Testing Reactor (JMTR). Crack growth rate and its electrochemical corrosion potential (ECP) dependence are different between in-pile test and post irradiation examination (PIE), but these differences are not fully understood. The objectives of the present study are to understand the difference between in-pile and out-of-pile IASCC growth and to confirm the effectiveness of mitigation due to lowering ECP on in-pile crack growth rates. For in-pile crack growth tests, we have selected a large compact tension specimen such as 0.5T-CT because of validity of SCC growth test at a high stress intensity factor (K-value). For loading a 0.5T-CT specimen up to K - 30 MPa {radical}m, we have adopted a lever type loading unit for in-pile crack growth tests in the JMTR. In this report, an in-pile test plan for crack growth of irradiated SUS316L stainless steels under simulated BWR conditions in the JMTR and current status of development of in-pile crack growth test techniques are presented. (author)

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

Energy Technology Data Exchange (ETDEWEB)

Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2009-04-15

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than {approx}7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

Science.gov (United States)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Results of first catamnestical examinations of children of parents irradiated preconceptionally

International Nuclear Information System (INIS)

Herrmann, T.; Eberhardt, H.J.; Rupprecht, E.; Voigtmann, L.; Jochem, I.; Medizinische Akademie, Dresden

1976-01-01

Results of first preliminary examinations of 21 children whose parents were preconceptionally irradiated for tumor treatment are presented. Most striking variations have been found in connective tissue and skeleton. Carporadiograms proved to be a valuable means for examining minus variants in cases of normal clinical evidence. In the case of a child of an irradiated father malformations of the bones of hands and feet have been observed. There were two premature still-births out of three pregnancies in the case of a patient exposed to a high gonad dose. No obvious deviations could be observed in sense organs and the central nervous system. The sex ratio of children of irradiated women was significantly shifted in favour of girls. It is proposed to establish for the GDR a central file of data on descendants from persons irradiated preconceptionally. The present study may be considered as a model. Finally, the information to be given to radiotherapy patients in the generative age is dealt with. (author)

Results of first catamnestical examinations of children of parents irradiated preconceptionally

Energy Technology Data Exchange (ETDEWEB)

Herrmann, T; Eberhardt, H J; Rupprecht, E; Voigtmann, L; Jochem, I [Medizinische Akademie, Dresden (German Democratic Republic). Kinderklinik; Medizinische Akademie, Dresden (German Democratic Republic). Radiologische Klinik)

1976-12-01

Results of first preliminary examinations of 21 children whose parents were preconceptionally irradiated for tumor treatment are presented. Most striking variations have been found in connective tissue and skeleton. Carporadiograms proved to be a valuable means for examining minus variants in cases of normal clinical evidence. In the case of a child of an irradiated father malformations of the bones of hands and feet have been observed. There were two premature still-births out of three pregnancies in the case of a patient exposed to a high gonad dose. No obvious deviations could be observed in sense organs and the central nervous system. The sex ratio of children of irradiated women was significantly shifted in favour of girls. It is proposed to establish for the GDR a central file of data on descendants from persons irradiated preconceptionally. The present study may be considered as a model. Finally, the information to be given to radiotherapy patients in the generative age is dealt with.

Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Khericha, S.T.

2002-06-30

This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

PIE of nuclear grade SiC/SiC flexural coupons irradiated to 10 dpa at LWR temperature

Energy Technology Data Exchange (ETDEWEB)

Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-03-01

Silicon carbide fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230–340°C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials are chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC)-coated Hi-NicalonTM Type-S (HNS), TyrannoTM SA3 (SA3), and SCS-Ultra^TM (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young’s modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young’s moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

Examination of irradiated fuel elements using gamma scanning technique

International Nuclear Information System (INIS)

Ichim, O.; Mincu, M.; Man, I.; Stanica, M.

2016-01-01

The purpose of this paper is to validate the gamma scanning technique used to calculate the activity of gamma fission products from CANDU/TRIGA irradiated fuel elements. After a short presentation of the equipments used and their characteristics, the paper describes the calibration technique for the devices and how computed tomography reconstruction is done. Following the previously mentioned steps is possible to obtain the axial and radial profiles and the computed tomography reconstruction for calibration sources and for the irradiated fuel elements. The results are used to validate the gamma scanning techniques as a non-destructive examination method. The gamma scanning techniques will be used to: identify the fission products in the irradiated CANDU/TRIGA fuel elements, construct the axial and radial distributions of fission products, get the distribution in cross section through computed tomography reconstruction, and determine the nuclei number and the fission products activity of the irradiated CANDU/TRIGA fuel elements. (authors)

Tratamiento del pie diabético en pacientes ancianos con Heberprot-P

Directory of Open Access Journals (Sweden)

José Luís Solís Licea

2015-12-01

Full Text Available El pie diabético constituye la complicación crónica de mayor morbimortalidad de la diabetes mellitus, significando un problema de salud internacional, del cual no escapa nuestro país y nuestro sistema de salud. Se realizó un estudio de tipo clínico observacional y descriptivo, con el objetivo de evaluar el resultado del tratamiento con Heberprot-P en pacientes con úlceras del pie diabético, en ancianos ingresados en el servicio de Angiología del Hospital General Docente “Dr. Ernesto Guevara de la Serna”, del municipio de Las Tunas, provincia de Las Tunas, en el período comprendido entre el 1ro de enero de 2010 al 31 de diciembre de 2011. Se estudiaron los pacientes según las variables de edad, localización de la lesión, resultado del tratamiento según la clasificación de Wagner y reacciones adversas presentadas. Con este estudio se pretende poner en manos del personal de la salud un material de consulta, que permita evaluar el tratamiento con Heberprot-P del pie diabético en el anciano.

Pie plano astrágalo vertical asociado a sinovitis. Fases de tratamiento

OpenAIRE

Planell i Mas, Elena de; Bonilla Toyos, Elvira

2008-01-01

El pie plano astrágalo vertical congénito es una deformidad poco frecuente que se puede encontrar de forma aislada, dentro de un cuadro polimalformativo o en relación con afectaciones neurológicas. Se caracteriza por una luxación del escafoides por encima del astrágalo,lo que provoca un pie rígido o de dificil manipulación. En este artículo los autores describen un caso clínico de un paciente que acude a consulta por presentar una inflamación dolorosa en la zona de la cabeza del astrágalo del...

Evaluation of neutron irradiation effect on SCC crack growth behaviour for austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Austenitic stainless steels are widely used as structural components in reactor pressure vessel internals because of their high strength, ductility, and fracture toughness. However, exposure to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the steels. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term plant operation in high temperature water environments is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) in Type 304 stainless steel in BWR environment. JNES had been conducting IASCC project during the JFY (2000) - JFY (2008) period, and prepared an engineering database on IASCC. However, the data of Crack Growth Rate (CGR) below the critical fluence level are not sufficient. So, this project was initiated to obtain the CGR data below the critical fluence level. Test specimens have been irradiated in the Halden reactor, operating by the OECD Halden Reactor Project, and the post irradiation examination (PIE) will be conducted from JFY (2011) to JFY (2013), finally the modified IASCC guide will be prepared in JFY (2013). (author)

Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

2008-01-01

U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation – Non-destructive analysis of the AFIP-1 fuel plates

Energy Technology Data Exchange (ETDEWEB)

Wachs, D.M., E-mail: daniel.wachs@inl.gov [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Robinson, A.B.; Rice, F.J. [Idaho National Laboratory, Characterization and Advanced PIE Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Kraft, N.C.; Taylor, S.C. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Lillo, M. [Idaho National Laboratory, Nuclear Systems Design and Analysis Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Woolstenhulme, N.; Roth, G.A. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008–2009. The irradiation conditions were: ∼250 W/cm{sup 2} peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm{sup 3} peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Post irradiation examinations cooperation and worldwide utilization of facilities

International Nuclear Information System (INIS)

Karlsson, Mikael

2009-01-01

Status of post irradiation examinations in Studsvik's facilities, cooperation and worldwide utilization of facilities, was described. Studsvik cooperate with irradiation facilities, as Halden, CEA and JAEA, as well as other hot cell facilities (examples, PSI, ITU and NFD) universities (example, the Royal Institute of Technology in Sweden) in order to be able to provide everything asked for by the nuclear community. Worldwide cooperation for effective use of expensive and highly specialized facilities is important, and the necessity of cooperation will be more and more recognized in the future. (author)

International Workshop on Characterization and PIE Needs for Fundamental Understanding of Fuels Performance and Safety

Energy Technology Data Exchange (ETDEWEB)

Not Listed

2011-12-01

The International Workshop on Characterization and PIE Needs to Support Science-Based Development of Innovative Fuels was held June 16-17, 2011, in Paris, France. The Organization for Economic Co-operation and Development (OECD), Nuclear Energy Agency (NEA) Working Party on the Fuel Cycle (WPFC) sponsored the workshop to identify gaps in global capabilities that need to be filled to meet projected needs in the 21st century. First and foremost, the workshop brought nine countries and associated international organizations, together in support of common needs for nuclear fuels and materials testing, characterization, PIE, and modeling capabilities. Finland, France, Germany, Republic of Korea, Russian Federation, Sweden, Switzerland, United Kingdom, United States of America, IAEA, and ITU (on behalf of European Union Joint Research Centers) discussed issues and opportunities for future technical advancements and collaborations. Second, the presentations provided a base level of understanding of current international capabilities. Three main categories were covered: (1) status of facilities and near term plans, (2) PIE needs from fuels engineering and material science perspectives, and (3) novel PIE techniques being developed to meet the needs. The International presentations provided valuable data consistent with the outcome of the National Workshop held in March 2011. Finally, the panel discussion on 21st century PIE capabilities, created a unified approach for future collaborations. In conclusion, (1) existing capabilities are not sufficient to meet the needs of a science-based approach, (2) safety issues and fuels behavior during abnormal conditions will receive more focus post-Fukushima; therefore we need to adopt our techniques to those issues, and (3) International collaboration is needed in the areas of codes and standards development for the new techniques.

Status of fuel irradiation tests in HANARO

International Nuclear Information System (INIS)

Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

1999-01-01

Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

Summary of the Minor Actinide-bearing MOX AFC-2C and -2D Irradiations

International Nuclear Information System (INIS)

McClellan, Kenneth; Chichester, Heather; Hayes, Steve; Voit, Stewart

2013-01-01

Summary of AFC-2C and AFC-2D tests: • AFC-2C and 2D, 1st MOX experiments in FCRD, were irradiated in ATR; • Initial results indicate performance of experimental MA-MOX fuels are similar to standard FR MOX fuels; • Cd-shrouded ATR experiment assembly and 235 U enrichment produce prototypic fast reactor power and temperature profiles leading to classic MOX zone restructuring; • Baseline postirradiation examinations have been completed for AFC-2C MOX and MA-MOX fuels; • Future work includes: – PIE of AFC-2D; – compare results to prototypic MOX fuel performance; – electron microscopy for microstructure and constituent distribution; – advanced NDE on saved pins

Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

International Nuclear Information System (INIS)

Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

1982-01-01

The document reports in detail the fuel pin fabrication data and describes the irradiation conditions and history. All the relevant results of the non-destructive and destructive post-irradiation examinations are reported. They include: visual inspection and chemical analysis of crud; length and diameter measurements; neutron radiography and gamma scanning; juncture tests and fission gas analysis (including residual gas in fuel samples); microscopy and alpha + beta/gamma autoradiography; microprobe investigations; burn-up and isotopic analysis; and hydrogen analysis in clad. The data and observations obtained are discussed in detail and conclusions are given. The irradiation and post-irradiation examinations of the R-109 pins have shown the safe, pre-calculable performance of LWR fuel pins containing mixed-oxide sphere-pac fuel with the fissile material mainly present in the large spheres

Advanced Post-Irradiation Examination Capabilities Alternatives Analysis Report

Energy Technology Data Exchange (ETDEWEB)

Jeff Bryan; Bill Landman; Porter Hill

2012-12-01

An alternatives analysis was performed for the Advanced Post-Irradiation Capabilities (APIEC) project in accordance with the U.S. Department of Energy (DOE) Order DOE O 413.3B, “Program and Project Management for the Acquisition of Capital Assets”. The Alternatives Analysis considered six major alternatives: ? No Action ? Modify Existing DOE Facilities – capabilities distributed among multiple locations ? Modify Existing DOE Facilities – capabilities consolidated at a few locations ? Construct New Facility ? Commercial Partnership ? International Partnerships Based on the alternatives analysis documented herein, it is recommended to DOE that the advanced post-irradiation examination capabilities be provided by a new facility constructed at the Materials and Fuels Complex at the Idaho National Laboratory.

Alteraciones biomecánicas del antepie en el pie zambo y su tratamiento

OpenAIRE

Oller Asensio, Antonio; Prats Climent, Baldiri

1991-01-01

Definición y recuerdo anátomo-fisio-patológico del pie zambo y de las alteraciones funcionales más frecuentes que se encuentran asociadas a esta patología. Se define la metodología básica de exploración física del paciente y estudio radiológico con la determinación de las tangentes tibio-calcáneas. Desarrollo de los tratamientos generales, ortopodológicos, fisioterapéuticos y quirúrgicos. Presentación de tres casos clínicos afectos de alteraciones biomecánicas por pie zambo con su tratamiento...

RERTR-12 Post-irradiation Examination Summary Report

Energy Technology Data Exchange (ETDEWEB)

Rice, Francine [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williams, Walter [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robinson, Adam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Meyer, Mitch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabin, Barry [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-02-01

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding.

Uso del método de Ilizarov en pie bot pediátrico. [Use of the Ilizarov method in pediatric clubfoot

Directory of Open Access Journals (Sweden)

Silvana Fiscina

2013-05-01

Full Text Available Introducción El objetivo de tratamiento del pie Bot está enfocado en lograr un pie plantígrado, indoloro y  que permita el uso del calzado habitual. El objetivo del trabajo fue medir la incidencia de recidiva luego de obtenida la corrección de la deformidad en pacientes con pie bot tratados con tutor de Ilizarov, y  analizar los factores pronósticos asociados. Material y Método Cohorte retrospectiva de factores pronóstico. Se trató con este método a 48 pacientes (70 pies desde 1988. Se excluyeron los pacientes con falta de corrección de la deformidad en el momento de retirar el tutor (deformidad residual o corrección incompleta. La variable de estudio fue la recidiva. Resultados El 77% de los 70 pies eran idiopáticos. Solo 18 pies no habían sido tratados. Todos eran pacien- tes pediátricos. El promedio de edad en el momento de la cirugía era de 7,9 años. Se incluyeron tres grupos de pacientes: tratados con Ilizarov solo (40,2%, en combi- nación con cirugía de partes blandas (38,5% y asociado con osteotomías (21,3%. El seguimiento promedio posoperatorio fue de 10,6 años (de 5,5 a 22. En el 71% de los pies, los resultados fueron pobres, la deformidad recidivó en un promedio de 38 meses, y fueron necesarias cirugías complementarias para lograr un pie plantígrado. Conclusión Presentamos una importante casuística con seguimiento alejado de pie bot recidivado o inveterado, con una alta recidiva de la deformidad (71%. No encontramos diferencias estadísticamente significativas entre los subgrupos estudiados (edad, diagnóstico, trata- mientos previos y procedimientos asociados. Actualmente no consideramos a esta técnica de primera elección, pues nuestros resultados fueron poco alentadores.

Axial gap formation in P.W.R. fuel pins

International Nuclear Information System (INIS)

Roberts, G.; Jones, K.W.

1978-07-01

The potential mechanisms of axial gap formation in PWR fuel pins are examined analytically and also using evidence from post-irradiation examination (p.i.e.) investigation. It is concluded that fuel and cladding cannot remain in contact during densification and so the settling of of the fuel stack, which forms the gaps, must be prevented by such things as asperities in the cladding, fuel chips or tilted pellets. Examples from the p.i.e. examination programme are used to support this conclusion. (author)

49 CFR 40.403 - Must a service agent notify its clients when the Department issues a PIE?

Science.gov (United States)

2010-10-01

... 49 Transportation 1 2010-10-01 2010-10-01 false Must a service agent notify its clients when the... Must a service agent notify its clients when the Department issues a PIE? (a) As a service agent, if... clients, in writing, about the issuance, scope, duration, and effect of the PIE. You may meet this...

Design and fabrication report on capsule (11M 19K for out of pile test) for irradiation testing of research reactor materials at HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, B.G.; Yang, S.W.; Park, S.J.; Shim, K.T.; Choo, K.N.; Oh, J.M.; Lee, B.C.; Choi, M.H.; Kim, D.J.; Kim, J.M.; Kang, S.H.; Chun, Y.B.; Kim, T.K.; Jeong, Y.H.

2012-05-15

As a part of the research reactor development project with a plate type fuel, the irradiation tests of graphite (Gr), beryllium (Be), and zircaloy 4 materials using the capsule have been investigating to obtain the mechanical characteristics such as an irradiation growth, hardness, swelling and tensile strength at the temperature below 100 .deg. C and the 30 MW reactor power. Then, A capsule to be able to irradiate materials(graphite, Be, zircaloy 4) under 100 .deg. C at the HANARO was designed and fabricated. After performing out of pile testing in single channel test loop by using the capsule, the final design of the capsules to be irradiated in CT and IR2 test hole of HANARO was approved, and 2 sets of capsule were fabricated. These capsules will be loaded in CT and IR2 test hole of HANARO, and be started the irradiation from the end of June, 2012. After performing the irradiation testing of 2 sets of capsule, PIE (Post Irradiation Examination) on irradiated specimens (Gr, Be, and zircaloy 4) will be carry out in IMEF (Irradiated Material Examination Facility). So, the irradiation testing will be contributed to obtain the characteristic data induced neutron irradiation on Gr, Be, and zircaloy 4. And then, it is convinced that these data will be also contributed to obtain the license for JRTR (Jordan Research and Training Reactor) and new research reactor in Korea, and export research reactors.

Magnetic Resonance Imaging: A Tool for Pork Pie Development

Science.gov (United States)

Gaunt, Adam P.; Morris, Robert H.; Newton, Michael I.

2013-01-01

The traditional British pork pie consists of roughly chopped pork cooked in a hot water pastry crust. Due to shrinkage of the meat during cooking, the gap formed around the meat is usually sealed using a gelatin based jelly to exclude air and thus help to preserve the pie. The properties of the jelly are such that it will ingress into the pastry crust causing undesirable softening. The jelly is traditionally produced by simmering pig trotters with seasoning for several hours. In this work we demonstrate the potential of magnetic resonance imaging (MRI) as a tool for investigating the conditions required for producing jellies with different properties and present two examples of this use. Firstly we demonstrate that MRI can determine the ability of water to diffuse through the jelly which is critical in minimizing the amount of moisture moving from the jelly to the crust. Secondly, the impact of jelly temperature on the penetration length into the crust is investigated. These examples highlight the power of MRI as a tool for food assessment. PMID:28239124

Dresden 1 plutonium recycle program. Final report

International Nuclear Information System (INIS)

Bresnick, S.D.

1980-01-01

This is the final report on the Dresden 1 Plutonium Recycle Demonstration Program. It covers the work performed from July 1, 1978 to completion, which includes in-pool inspection of two fuel assemblies, removal of two fuel rods, and post-irradiation examination (PIE) of six fuel rods. Appendix A describes the inspection and rod removal operations, and Appendix B describes the PIE work

Clínica del pie valgo

OpenAIRE

Marcelino Reyes, Jesús

1998-01-01

Genericamente reconocemos al pie valgo, por las características siguientes: - desviación de la linea de Helbing ( retropié) - aplanamiento de la bóveda longitudinal interna - pronación de antepié. - aspecto de pié plano. Pero no podemos contentarnos con estas cuatro reglas,y por lo tanto hemos de pormenorizar y clasificar en función de los diferentes elementos que intervienen en la morfogenética de pié. Con lo cual también tendremos una visión mas amplia a la hora de poder aplicar un tratamie...

Microbiological evaluation of sweet pies marketed fairs in special city Goiânia-GO

Directory of Open Access Journals (Sweden)

Edmar Soares Nicolau

2014-02-01

Full Text Available The sweet pies are widely consumed in special fairs in Goiânia-GO city. In general they are marketed in inadequate way, without refrigeration and most of times without any protection against various types of contaminants, including microbial contamination that can cause food toxi-infections to consumers. This study aimed to evaluate the microbiological quality of pies sold in 105 stalls in 23 special fair city of Goiânia, and also the impact of training on the Good Manufacturing Practices (GMP in the quality of these products. The micro-organisms investigated were those required the by standard microbiological: Salmonella, fecal coliforms, Bacillus cereus and Staphylococcus coagulase positive. We also observed the temperature exposure of these foods on newsstands. We analyzed 259 samples in total, with 209 in the first stage, before the training of the fairground, and 50 in the second, after qualifying for the fair dealers in GMP. Pies sweets collected before training of fair dealers were considered unfit for consumption because it contains health indicators above the recommended limit (65.6% with fecal coliforms, 20.6% Staphylococcus coagulase positive and 16.8% with Bacillus cereus. In samples collected after training there was an improvement in the indices (28,0% contained coliforms, 14,0% Staphylococcus coagulase positive and 28,0% of Bacillus cereus, all above the limit set. None samples had Salmonella in both steps. It can be concluded that with the training of fair dealers there was an improvement in the microbiological quality of these products. These results reinforce the need for training of the fairground and the impact it can have on the quality of the products. The storage temperatures pies ranged from –5.3ºC to 28.7ºC in first step and –3.6ºC to 24.0ºC in the second. The presence of the refrigerated was not sufficient to maintain the optimal temperature (to 10 ° C for storage of pies.

Oxidative stress in pied flycatcher (Ficedula hypoleuca) nestlings from metal contaminated environments in northern Sweden

International Nuclear Information System (INIS)

Berglund, A.M.M.; Sturve, J.; Foerlin, L.; Nyholm, N.E.I.

2007-01-01

Metals have been shown to induce oxidative stress in animals. One of the most metal polluted terrestrial environments in Sweden is the surroundings of a sulfide ore smelter plant located in the northern part of the country. Pied flycatcher nestlings (Ficedula hypoleuca) that grew up close to the industry had accumulated amounts of arsenic, cadmium, mercury, lead, iron and zinc in their liver tissue. The aim of this study was to investigate if pied flycatcher nestlings in the pollution gradient of the industry were affected by oxidative stress using antioxidant molecules and enzyme activities. The antioxidant assays were also evaluated in search for useful biomarkers in pied flycatchers. This study indicated that nestlings in metal contaminated areas showed signs of oxidative stress evidenced by up regulated hepatic antioxidant defense given as increased glutathione reductase (GR) and catalase (CAT) activities and slightly but not significantly elevated lipid peroxidation and glutathione-S-transferase (GST) activities. Stepwise linear regression indicated that lipid peroxidation and CAT activities were influenced mostly by iron, but iron and lead influenced the CAT activity to a higher degree. Positive relationships were found between GST and lead as well as GR activities and cadmium. We conclude that GR, CAT, GST activities and lipid peroxidation levels may function as useful biomarkers for oxidative stress in free-living pied flycatcher nestlings exposed to metal contaminated environments

Tensile mechanical properties of a stainless steel irradiated up to 19 dpa in the Swiss spallation neutron source

Energy Technology Data Exchange (ETDEWEB)

Saito, Shigeru, E-mail: saito.shigeru@jaea.go.jp [JAEA, J-PARC Center, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kikuchi, Kenji [Ibaraki Univ., iFRC, Tokai-mura, Ibaraki-ken 319-1106 (Japan); Hamaguchi, Dai [JAEA, J-PARC Center, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Usami, Kouji; Endo, Shinya; Ono, Katsuto; Matsui, Hiroki [JAEA, Dept. of Hot Laboratories, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kawai, Masayoshi [KEK, Tsukuba-shi, Ibaraki-ken 305-0801 (Japan); Dai, Yong [PSI, Spallation Source Division, Villigen PSI (Switzerland)

2012-12-15

To evaluate the lifetime of the beam window of an accelerator-driven transmutation system (ADS), post irradiation examination (PIE) of the STIP (SINQ target irradiation program, SINQ; Swiss spallation neutron source) specimens was carried out. The specimens tested in this study were made from the austenitic steel Japan primary candidate alloy (JPCA). The specimens were irradiated at SINQ Target 4 (STIP-II) with high-energy protons and spallation neutrons. The irradiation conditions were as follows: the proton energy was 580 MeV, irradiation temperatures ranged from 100 to 430 Degree-Sign C, and displacement damage levels ranged from 7.1 to 19.5 dpa. Tensile tests were performed in air at room temperature (RT), 250 Degree-Sign C and 350 Degree-Sign C. Fracture surface observation after the tests was done by Scanning electron microscope (SEM). Results of the tensile tests performed at R.T. showed the extra hardening of JPCA at higher dose compared to the fission neutron irradiated data. At the higher temperatures, 250 Degree-Sign C and 350 Degree-Sign C, the extra hardening was not observed. Degradation of ductility bottomed around 10 dpa, and specimens kept their ductility until 19.5 dpa. All specimens fractured in ductile manner.

Report of Post Irradiation Examination for Dry Process Fuel

International Nuclear Information System (INIS)

Par, Jang Jin; Jung, I. H.; Kang, K. H.; Moon, J. S.; Lee, C. R.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Yoo, B. O.; Jung, Y. H.; Choo, Y. S.

2006-08-01

The spent PWR fuel typically contains 0.9 wt.% of fissile uranium and 0.6 wt.% of fissile plutonium, which exceeds the natural uranium fissile content of 0.711 wt.%. The neutron economy of a CANDU reactor is sufficient to utilize the DUPIC fuel, even though the neutron-absorbing fission products contained in the spent PWR fuel were remained in the DUPIC fuel. The DUPIC fuel cycle offers advantages to the countries operating both the PWR and CANDU reactors, such as saving the natural uranium, reducing the spent fuel in both PWR and CANDU, and acquiring the extra energy by reuse of the PWR spent fuel. This report contains the results of post-irradiation examination of the DUPIC fuel irradiated four times at HANARO from May 2000 to August 2006 present except the first irradiation test of simulated DUPIC fuel at HANARO on August 1999

Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

International Nuclear Information System (INIS)

Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

2008-08-01

The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

The Perceived Invalidation of Emotion Scale (PIES): Development and psychometric properties of a novel measure of current emotion invalidation.

Science.gov (United States)

Zielinski, Melissa J; Veilleux, Jennifer C

2018-05-24

Emotion invalidation is theoretically and empirically associated with mental and physical health problems. However, existing measures of invalidation focus on past (e.g., childhood) invalidation and/or do not specifically emphasize invalidation of emotion. In this article, the authors articulate a clarified operational definition of emotion invalidation and use that definition as the foundation for development of a new measure of current perceived emotion invalidation across a series of five studies. Study 1 was a qualitative investigation of people's experiences with emotional invalidation from which we generated items. An initial item pool was vetted by expert reviewers in Study 2 and examined via exploratory factor analysis in Study 3 within both college student and online samples. The scale was reduced to 10 items via confirmatory factor analysis in Study 4, resulting in a brief but psychometrically promising measure, the Perceived Invalidation of Emotion Scale (PIES). A short-term longitudinal investigation (Study 5) revealed that PIES scores had strong test-retest reliability, and that greater perceived emotion invalidation was associated with greater emotion dysregulation, borderline features and symptoms of emotional distress. In addition, the PIES predicted changes in relational health and psychological health over a 1-month period. The current set of studies thus presents a psychometrically promising and practical measure of perceived emotion invalidation that can provide a foundation for future research in this burgeoning area. (PsycINFO Database Record (c) 2018 APA, all rights reserved).

Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

International Nuclear Information System (INIS)

Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; Maki, John T.

2014-01-01

The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of fission products silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of fission products from fuel compacts and fuel particles, and retention of fission products in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of these fission products was determined and compared to the PIE measurements. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of about two, corresponding to an over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of about 100. For intact particles, whose release is much lower, the over-prediction is by an average of about an order of magnitude, which could additionally be attributed to an over-estimated diffusivity in SiC by about 30%. The release of strontium from intact particles is also over-estimated by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Furthermore, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the

Fitness costs of polygyny in female Pied Flycatchers Ficedula hypoleuca

NARCIS (Netherlands)

Both, C.

2002-01-01

Variance in mating systems can only be understood if the fitness consequences of the available options are known for both sexes. In Pied Flycatchers Ficedula hypoleuca it has for long been suggested that females pairing with already mated males pay a fitness cost, and it is still debated why females

Prevención de úlceras en el pie del paciente diabético

Directory of Open Access Journals (Sweden)

Análida Pinilla Roa

2000-01-01

Full Text Available La Diabetes Mellitus tipo dos (1 es una de las enfermedades metabólicas más comunes dentro de la población adulta (mayor de 30 años, da lugar a múltiples complicaciones como insuficiencia renal, enfermedades cardiovasculares, ceguera, neuropatía periférica y amputaciones de los miembros inferiores. Vale la pena destacar el profundo impacto de la diabetes en los pies debido a la gran morbilidad y mortalidad que puede causar. Se sabe que los problemas en los pies de los diabéticos son el resultado de varios factores como son la neuropatía periférica, la vasculopatía, los trastornos de la inmunidad  y los traumatismos entre otros. Cada uno de estos factores contribuye a la formación de úlceras que no sanan fácilmente y en general se infectan causando un proceso séptico que requiere el tratamiento radical de la amputación. El objetivo de este artículo es concientizar al personal de salud sobre  la necesidad de la prevención de las úlceras en los pies de los pacientes diabéticos, mediante la educación para permitir una detección temprana y un tratamiento oportuno, disminuir la tasa de úlceras y por ende de amputaciones de los miembros inferiores. La educación al paciente y al personal de salud debe incluir: el reconocimiento de signos tempranos de ulceración en las áreas de mayor presión de los pies mediante un adecuado autoexamen diario, la asesoría para el uso de calzado protector, adecuados hábitos de aseo y corte de uñas, así como medidas preventivas para evitar el trauma.

[Characterization and nutritional value of a food artisan: the meat pie of Murcia].

Science.gov (United States)

Ruiz-Cano, Domingo; Pérez-Llamas, Francisca; López-Jiménez, José Ángel; González-Silvera, Daniel; Frutos, Maria José; Zamora, Salvador

2013-01-01

The main aims of this study are to describe the characteristics of the meat pie, a typical product of the regional gastronomy of Murcia and to determine its nutritional and energy values, fatty acid profile and fat quality. There were studied 24 samples of Murcia's meat pie from the six best-selling retail establishments in this Region (four units per establishment).The moisture, protein, fat, carbohydrates, fibre and minerals, contents and the energy value, fatty acid profile and fat quality were analyzed using the Official Analysis Methods of Foods. All analyses were performed by triplicate. The average weight of this product was 192.3 ± 11.8 g, with three differentiated parts (base, filling and pastry lid). All ingredients were natural raw materials: wheat flour, lard, ground beef, sliced boiled egg and chorizo, water and spices (salt, pepper, garlic, paprika and nutmeg). Most of its organoleptic attributes are due to the type and amount of fat or lard. The combination of the other ingredients and the particular formulation of the spices are also responsible of other attributes. Due to its protein content (11.0%), this meat pie can replace other meat dishes, and be incorporated into a balanced diet. However, it is necessary to take into account its fat and energy contents (17.3 g and 317 kcal/100 g, respectively). Unlike many common pastry products, it contains no trans fatty acids. The results of the research show that the studied product remains, at present, an artisan food, and offer reliable information that it is representative of the energy and nutritional values of the Murcia's meat pie, a typical product of the gastronomy of the Region of Murcia. Copyright © AULA MEDICA EDICIONES 2013. Published by AULA MEDICA. All rights reserved.

Valoración del pie plano en personas adultas con síndrome de Down y de su afectación en la vida diaria

OpenAIRE

Stafford, Frances

2017-01-01

Se realizó un estudio en 50 adultos con síndrome de Down, entre 18 y 40 años para analizar la incidencia e influencia del pie plano sobre el funcionamiento de la vida diaria. La incidencia fue del 70%, con grados diversos de pie plano. Hubo dolor en 4 personas, si bien era debido a otros problemas y no achacable al pie plano. En ningún caso la presencia de pie plano afectó al funcionamiento de la vida ordinaria.

Specific Heat Capacity of Alloy 690 for Simulating Neutron Irradiation

International Nuclear Information System (INIS)

Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Joo, Young Sun; Ahn, Sang Bok; Park, Jin Seok; Lee, Won Jae; Ryu, Woo Seok

2011-01-01

The KAERI(Korea Atomic Energy Research Institute) is developing new type of nuclear reactor, so called 'SMART'(System Integrated Modular Advanced Reactor) which has many features of small power and system integrated modular type. Alloy 690 was selected as the candidate material for the heat exchanger tube of the steam generator of SMART. The SMART R and D is now facing the stage of engineering verification and approval of standard design to apply to DEMO reactors. Therefore, the material performance under the relevant environment is required to be evaluated. The important material performance issues are mechanical properties i.e. (fracture toughness, tensile and hardness) and thermal properties i.e. (thermal diffusivity, specific heat capacity and thermal conductivity) for which the engineering database is necessary to design a steam generator. However, the neutron post irradiation characteristics of the alloy 690 are barely known. As a result, PIE(Post Irradiation Examination) of thermal properties are planed and performed successfully. But specific heat capacity measurement is not performed because of not having proper test system for irradiated materials. Therefore in order to verify the effect of neutron irradiation for alloy 690, simulation method is adopted. In general, high energy neutron bombardment in material bring about lattice defects i.e. void, pore and dislocation. Dominant factor to impact to heat capacity is mainly dislocation in material. Therefore, simulation of neutron irradiation is devised by material rolling method in order to make artificial dislocation in alloy 690 as same effect of neutron irradiation. After preparing test specimens, heat capacity measurements are performed and results are compared with rolled materials and un-rolled materials to verify the effect of neutron irradiation simulation. Main interest of simulation is that heat capacity value is changed by neutron irradiation

Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

2014-01-01

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

Post-irradiation examination of fuel elements of Tarapur Atomic Power Station (Report-I)

International Nuclear Information System (INIS)

Bahl, J.K.; Sah, D.N.; Chatterjee, S.; Sivaramkrishnan, K.S.

1979-01-01

Detailed post-irradiation examination of three initial load fuel elements of the Tarapur Atomic Power Station (TAPS) has been carried out. The causes of the element failures have been analysed. It was observed that almost 90% of the length of the elements exoerienced nodular corrosion. It has been estimated that nodular corrosion would seriously affect the wall thickness and surface temperature of higher rated elements. Lunar shaped fret marks have also been observed at some spacer grid locations in the elements. The depth of the largest fret mark was measured to be 16.9% clad wall thickness. Detailed metallographic examination of the clad and fuel in the three elements has been done. The temperatures at different structural regions of the fuel cross-sections have been estimated. The change in fuel density during irradiation has been evaluated by comparing the irradiated fuel diameter with the mean pellet design diameter. The performance of the end plug welds and spacer grid sites in the elements has been assessed. The burnup distribution along the length of the elements has been evaluated by gamma scanning. The redistribution of fission products in the fuel has been examined by gamma scanning and beta-gamma autoradiography. Mechanical properties of the irradiated cladding have been examined by ring tensile testing. (auth.)

The post irradiation examination of a sphere-pac (UPu)C fuel pin irradiated in the BR-2 reactor (MFBS 7 experiment)

International Nuclear Information System (INIS)

Smith, L.; Aerne, E.T.; Buergisser, B.; Flueckiger, U.; Hofer, R.; Petrik, F.

1979-09-01

A pin fuelled with Swiss made (UPu)C microspheres has been successfully irradiated to a peak burn-up of 6% fima in the Belgian BR2 Reactor. The pin, rated up to 95 kW/m, was intact after irradiation and exhibited a peak strain of just over 0.5%. The results of the post irradiation examination are reported. (Auth.)

AGR-2 Irradiated Test Train Preliminary Inspection and Disassembly First Look

Energy Technology Data Exchange (ETDEWEB)

Ploger, Scott [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowciz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

The AGR 2 irradiation experiment began in June 2010 and was completed in October 2013. The test train was shipped to the Materials and Fuels Complex in July 2014 for post-irradiation examination (PIE). The first PIE activities included nondestructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and their graphite fuel holders. Dimensional metrology was then performed on the compacts, graphite holders, and steel capsule shells. AGR 2 disassembly and metrology were performed with the same equipment used successfully on AGR 1 test train components. Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Disassembly of the AGR 2 test train and its capsules was conducted rapidly and efficiently by employing techniques refined during the AGR 1 disassembly campaign. Only one major difficulty was encountered while separating the test train into capsules when thermocouples (of larger diameter than used in AGR 1) and gas lines jammed inside the through tubes of the upper capsules, which required new tooling for extraction. Disassembly of individual capsules was straightforward with only a few minor complications. On the whole, AGR 2 capsule structural components appeared less embrittled than their AGR 1 counterparts. Compacts from AGR 2 Capsules 2, 3, 5, and 6 were in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated radial shrinkage between 0.8 to 1.7%, with the greatest shrinkage observed on Capsule 2 compacts that were irradiated at higher temperature. Length shrinkage ranged from 0.1 to 0.9%, with by far the lowest axial shrinkage on Capsule 3 compacts�

Nesting ecology of Pied Avocet Recurvirostra avosetta in Sfax salina ...

African Journals Online (AJOL)

In this paper we use the results of a one-year monitoring of nests in Sfax salina to provide information on its nesting parameters, in particular nesting phenology, colony size and hatching success. Our results show that Pied Avocets formed dense colonies at the beginning of the nesting season, but colony size decreased as ...

Microstructural examination of V-(4-5%) Cr-(4-5%)Ti irradiated in X530

Energy Technology Data Exchange (ETDEWEB)

Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States); Chung, H.M. [Argonne Natinonal Lab., IL (United States)

1997-08-01

Microstructural examination results are reported for two heats of V-(4-5%)Cr-(4-5%)Ti irradiated in the X530 experiment to {approximately}400{degrees}C to provide an understanding of the microstructural evolution that may be associated with degradation of mechanical properties. Fine precipitates were observed in high density intermixed with small defect clusters for all conditions examined following the irradiation. The irradiation-induced precipitation does not appear to be affected by preirradiation heat treatment at 950-1125{degrees}C. There was no evidence for a significant density of large (diameter >10 nm) dislocation loops or network dislocations.

Post-Irradiation Examination of Array Targets - Part I

Energy Technology Data Exchange (ETDEWEB)

Icenhour, A.S.

2004-01-23

During FY 2001, two arrays, each containing seven neptunium-loaded targets, were irradiated at the Advanced Test Reactor in Idaho to examine the influence of multi-target self-shielding on {sup 236}Pu content and to evaluate fission product release data. One array consisted of seven targets that contained 10 vol% NpO{sub 2} pellets, while the other array consisted of seven targets that contained 20 vol % NpO{sub 2} pellets. The arrays were located in the same irradiation facility but were axially separated to minimize the influence of one array on the other. Each target also contained a dosimeter package, which consisted of a small NpO{sub 2} wire that was inside a vanadium container. After completion of irradiation and shipment back to the Oak Ridge National Laboratory, nine of the targets (four from the 10 vol% array and five from the 20 vol% array) were punctured for pressure measurement and measurement of {sup 85}Kr. These nine targets and the associated dosimeters were then chemically processed to measure the residual neptunium, total plutonium production, {sup 238}Pu production, and {sup 236}Pu concentration at discharge. The amount and isotopic composition of fission products were also measured. This report provides the results of the processing and analysis of the nine targets.

Fractal Pied de Poule (houndstooth) Collection SS'15 : Parka and Jacket

NARCIS (Netherlands)

2015-01-01

Fractal Pied de Poule (houndstooth) Spring/Summer '15 is a collection consisting of a body, a jacket and a parka. The last two will be shown here. Algorithms, new materials, digital prototyping, drapability, tessellations and fractals are recurring themes in our projects. The body, the jacket and

Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation

Energy Technology Data Exchange (ETDEWEB)

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

2016-11-01

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory’s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry.

Annual report of JMTR, 1993

International Nuclear Information System (INIS)

Yamazaki, Hiroshi; Komori, Yoshihiro; Nakayama, Fusao

1994-12-01

In FY 1993 (April 1993 to March 1994), the JMTR was operated from the 106th cycle to the former half of the 109th cycle, and the JMTR was fully converted to the LEU fuels in the 108th cycle. Various capsules and rabbits were irradiated for nuclear fuels and materials research and for radioisotope production, and the related PIEs were carried out. R and D works on irradiation technology and PIE technology have been extensively continued as follows:Capsules have been developed for fusion materials irradiation experiments and fundamental research ; Re-instrumentation technique, remote control crack propagation testing machine, examination technique with miniaturized specimens and underwater cutting machine for high radioactive components have been developed;Irradiation tests on blanket materials has been intensively carried out. (author)

Annual report on operation, utilization and technical development of hot laboratories. From April 1, 2001 to March 31, 2002

International Nuclear Information System (INIS)

2003-01-01

This is an annual report in a fiscal year of 2001 that describes activities of the Reactor Fuel Examination Facility (RFEF), the Waste Safety Testing Facility (WASTEF), and the Research Hot Laboratory (RHL) in the Department of Hot laboratories. In RFEF, PIEs including destructive and nondestructive tests were performed on a BWR fuel assembly and/or its fuel rod irradiated in the Fukushima-2 Nuclear Power Station Unit-1 and a fuel assembly with UO 2 -Gd 2 O 3 and mixed oxide (MOX) fuel pellets for Japan Nuclear Cycle Development Institute. In addition, 34 fuel assemblies irradiated in the nuclear ship ''Mutsu'' were conveyed from Mutsu Establishment, and re-assembly and PIEs for the assemblies were carried out. In WASTEF, tests for evaluating barrier performance in terms of disposal of waste, high temperature tests for evaluating stable on TRansUraniums (TRU) nitrides, leaching tests on Rock-like OXide (ROX) fuels were performed. The slow Strain Rate Tests (SSRT) apparatuses were installed for investigation of Irradiation-Assisted Stress Corrosion Cracking (IASCC) on light water structural materials, and characterization tests for the apparatus were performed. In RHL, PIEs for light water reactor materials, fusion materials, and target materials of Proton Accelerator Facilities were carried out for laboratories in Japan Atomic Energy Research Institute. PIEs for zirconium alloys for ultra-high burn-up irradiated in the Kashiwazaki Nuclear Power Station Unit-5 were also performed. In order to investigate roots cause of pipe rupture in Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company, several examinations including SEM observation, EPMA, and Vickers hardness test were performed in those three facilities. The data from the examinations greatly contribute to clarify roots cause of the pipe rupture. (author)

Annual report on operation, utilization and technical development of hot laboratories. From April 1, 2001 to March 31, 2002

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-01-01

This is an annual report in a fiscal year of 2001 that describes activities of the Reactor Fuel Examination Facility (RFEF), the Waste Safety Testing Facility (WASTEF), and the Research Hot Laboratory (RHL) in the Department of Hot laboratories. In RFEF, PIEs including destructive and nondestructive tests were performed on a BWR fuel assembly and/or its fuel rod irradiated in the Fukushima-2 Nuclear Power Station Unit-1 and a fuel assembly with UO{sub 2}-Gd{sub 2}O{sub 3} and mixed oxide (MOX) fuel pellets for Japan Nuclear Cycle Development Institute. In addition, 34 fuel assemblies irradiated in the nuclear ship ''Mutsu'' were conveyed from Mutsu Establishment, and re-assembly and PIEs for the assemblies were carried out. In WASTEF, tests for evaluating barrier performance in terms of disposal of waste, high temperature tests for evaluating stable on TRansUraniums (TRU) nitrides, leaching tests on Rock-like OXide (ROX) fuels were performed. The slow Strain Rate Tests (SSRT) apparatuses were installed for investigation of Irradiation-Assisted Stress Corrosion Cracking (IASCC) on light water structural materials, and characterization tests for the apparatus were performed. In RHL, PIEs for light water reactor materials, fusion materials, and target materials of Proton Accelerator Facilities were carried out for laboratories in Japan Atomic Energy Research Institute. PIEs for zirconium alloys for ultra-high burn-up irradiated in the Kashiwazaki Nuclear Power Station Unit-5 were also performed. In order to investigate roots cause of pipe rupture in Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company, several examinations including SEM observation, EPMA, and Vickers hardness test were performed in those three facilities. The data from the examinations greatly contribute to clarify roots cause of the pipe rupture. (author)

Examination of the creep behaviour of ceramic fuel elements under neutron irradiation

International Nuclear Information System (INIS)

Brucklacher, D.

1978-01-01

This paper examines the creeping of UO 2 , UO 2 -PuO 2 and UN under neutron irradiation. It starts with the experimental results about the relation between the thermal creep rate and the load, the temperature, as well as characteristic material values, stoichiometry, grain size and porosity. These correlation are first qualitatively discussed and then compared with the statements of actual quantitative equations. From the models and theories on which these equations are based a modified Nabarro-Heering-equation results for the correlation between the creep rate of ceramic fuels, stress, temperature and the fission rate. In the experimental part of the examination, length-changes of creep samples of UO 2 , (U,Pu)O 2 and UN were measured in specially developed irradiation creep casings in different reactors. The measuring data were corrected and evaluated considering the thermal expansion effects, irregular temperature distribution and swelling effects in such a way that the dependences of the creep rate of UO 2 , UO 2 -PuO 2 and UN under irradiation on stress, temperature, fission rate, burn-up and porosity is obtained. It shows that creeping of fuels under irradiation at high temperatures is equivalent to thermally activated creeping, while at low temperature the creep rate induced by irradiation is much higher than the condition without irradiation. The increment of oxidic nuclear fuels is greater than in UN, the stress dependence on low burn-up is proportional in both cases, and the influence of temperature is quite small. (orig.) [de

Moult, pied plumage and relationships within the genus of the Black ...

African Journals Online (AJOL)

The pied plumage of the adult Black Sparrowhawk is rather exceptional in the genus Accipiter and it could be explained by functionality or by phylogenetic relationships. The moult pattern of museum specimens is presented, supplementing information from captive birds. The post-juvenile moulting sequence is similar to that ...

Irradiation and performance evaluation of DUPIC fuel

International Nuclear Information System (INIS)

Bae, Ki Kwang; Yang, M. S.; Song, K. C.

2000-05-01

The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis

Irradiation and performance evaluation of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Bae, Ki Kwang; Yang, M S; Song, K C [and others

2000-05-01

The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis.

Spatial distribution of breeding Pied Flycatchers Ficedula hypoleuca in respect to their natal sites

Directory of Open Access Journals (Sweden)

Sokolov, L.

2004-06-01

Full Text Available Study of philopatry and dispersal of pied flycatchers Ficedula hypoleuca was launched on the Courish Spit (SE Baltic in 1981. Since then, ca. 9,000 nestlings were ringed at different sites in the Russian part of the Courish Spit. A total of 557 individuals ringed as pulli were recaptured in subsequent seasons in the study area. Both males and females are more often recaptured in the plots where they were ringed than in other plots. These results were interpreted in the framework of the hypothesis forwarded by Löhrl (1959 and supported by Berndt & Winkel (1979. These authors suggested that cavity nesters (pied flycatchers and collared flycatchers F. albicollis imprint their future local breeding area during the period of postfledging exploration. Birds that survive until the next spring, return to these imprinted areas to breed. A similar study done by Sokolov et al. (1984 on the Courish Spit in an open nesting species, the chaffinch Fringilla coelebs, confirmed this finding. We assumed that juvenile pied flycatchers disperse for varying distances during their postfledging movements and imprint a local area, some 1–5 kilometres in diameter. This area is the goal of their migration next spring. It is suggested that in spring, yearlings are non–randomly distributed in respect to the area they have imprinted as juveniles. Recently, Vysotsky (2000, 2001 re–analysed the same data on philopatry of pied flycatchers on the Courish Spit and forwarded an alternative hypothesis. He suggests that juveniles, both males and females, do not imprint any local area during the postfledging period, but are distributed randomly across the area of several dozens of kilometres in spring. Vysotsky was able to show that distribution of distances of natal dispersal did not differ from the random pattern the study plot which was an 8.5 km long line of nest boxes along the Courish Spit. The aim of this study was to test these two alternative hypotheses. To do so, we

Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

International Nuclear Information System (INIS)

Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H.

2001-05-01

The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project T he Nuclear Fuel Material Development of Research Reactor . And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,

Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

Energy Technology Data Exchange (ETDEWEB)

Paul Demkowicz; Scott Ploger; John Hunn

2012-05-01

The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

International Nuclear Information System (INIS)

Demkowicz, Paul; Ploger, Scott; Hunn, John

2012-01-01

The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

Active basement uplift as seen with cosmogenic lenses: the Sierra Pie de Palo case (Western Argentina)

Science.gov (United States)

Siame, L. L.; Sébrier, M.; Costa, C. H.; Ahumada, E. A.; Bellier, O.

2013-12-01

The Andean foreland of western Argentina (28°S-33°S) corresponds to retroarc deformations associated with the ongoing flat subduction of the Nazca plate beneath the South American lithosphere, and associated with high levels of seismic activity and crustal active faulting. To improve earthquake source identification and characterization in the San Juan region, data from seismology, structural geology and quantitative geomorphology can be integrated and combined to provide a seismotectonic model. In this model, the Andean back-arc of western Argentina has to be regarded as an obliquely converging foreland where Plio-Quaternary deformations are partitioned between strike-slip and thrust motions that are localized on the E-verging, thin-skinned Argentine Precordillera, and the W-verging thick-skinned Sierras Pampeanas, respectively. In this domain, the Sierra Pie de Palo is a key structure playing a major role in the partitioning of the Plio-Quaternary deformations. Located in the westernmost Sierras Pampeanas, the Sierra Pie de Palo forms a NNE striking, 80 km-long and 35-40 km-wide, ellipsoid range that reaches elevation as high as 3162 m. This mountain range is an actively growing basement fold associated with a high level of seismic activity (e.g., the November 23, 1977, Caucete, Mw 7.4 earthquake). To evaluate the degree of tectonic activity around the Sierra Pie de Palo, we combined a detailed morphometric analysis of the topography together with in situ-produced cosmogenic 10Be concentrations measured in (1) bedrock outcrops corresponding to the exhumed erosional regional surface, (2) surface boulders abandoned on alluvial fans deformed by active faults, and (3) in fluvial sediments sampled at the outlets of selected watersheds that drains out from the Sierra Pie de Palo. All together, our results allows: (1) assessing quantitative constraints on the rate of tectonic and denudation processes that are responsible for the active growth and erosion of the Sierra

Giant Paperclip Necklaces, Soup-Can Rings and Cherry-Pie Hats

Science.gov (United States)

Winters, Laurel A.

2011-01-01

In this article, the author describes an art project inspired by the wearable sculpture art created by artist Marjorie Schick. Students used wallpaper paste and newspapers to create papier-mache for a mountain hat, a cherry-pie mask/hat, a "dress" shoe and a Cubistic mask. Cardboard was used in many of these things, in addition to being used as…

Data on post irradiation experiments of heat resistant ceramic composite materials. PIE for 97M-13A

Energy Technology Data Exchange (ETDEWEB)

Baba, Shin-ichi; Ishihara, Masahiro; Souzawa, Shizuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Sekino, Hajime [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

The research on the radiation damage mechanism of heat resistant ceramic composite materials is one of the research subjects of the innovative basic research in the field of high temperature engineering, using the High Temperature engineering Test Reactor (HTTR). Three series of irradiation tests on the heat resistant ceramic composite materials, first to third irradiation test program, were carried out using the Japan Material Testing Reactor (JMTR). This is a summary report on the first irradiation test program; irradiation induced dimensional change, thermal expansion coefficient, X-ray diffraction and {gamma}-ray spectrum are reported. (author)

Multi-MW accelerator target material properties under proton irradiation at Brookhaven National Laboratory linear isotope producer

Science.gov (United States)

Simos, N.; Ludewig, H.; Kirk, H.; Dooryhee, E.; Ghose, S.; Zhong, Z.; Zhong, H.; Makimura, S.; Yoshimura, K.; Bennett, J. R. J.; Kotsinas, G.; Kotsina, Z.; McDonald, K. T.

2018-05-01

The effects of proton beams irradiating materials considered for targets in high-power accelerator experiments have been studied using the Brookhaven National Laboratory's (BNL) 200 MeV proton linac. A wide array of materials and alloys covering a wide range of the atomic number (Z) are being scoped by the high-power accelerator community prompting the BNL studies to focus on materials representing each distinct range, i.e. low-Z, mid-Z and high-Z. The low range includes materials such as beryllium and graphite, the midrange alloys such as Ti-6Al-4V, gum metal and super-Invar and finally the high-Z range pure tungsten and tantalum. Of interest in assessing proton irradiation effects are (a) changes in physiomechanical properties which are important in maintaining high-power target functionality, (b) identification of possible limits of proton flux or fluence above which certain materials cease to maintain integrity, (c) the role of material operating temperature in inducing or maintaining radiation damage reversal, and (d) phase stability and microstructural changes. The paper presents excerpt results deduced from macroscopic and microscopic post-irradiation evaluation (PIE) following several irradiation campaigns conducted at the BNL 200 MeV linac and specifically at the isotope producer beam-line/target station. The microscopic PIE relied on high energy x-ray diffraction at the BNL NSLS X17B1 and NSLS II XPD beam lines. The studies reveal the dramatic effects of irradiation on phase stability in several of the materials, changes in physical properties and ductility loss as well as thermally induced radiation damage reversal in graphite and alloys such as super-Invar.

Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

International Nuclear Information System (INIS)

Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

1981-04-01

Three fuel pin bundles, R-109/1, 2 and 3, were irradiated in a PWR loop in the HFR at Petten during respectively 131, 57 and 57 effective full power days at average powers of approximately 39 kW.m -1 and at peak powers of approximately 60 kW.m -1 . The results of the post-irradiation examinations of these fuel bundles are presented. (Auth.)

Post-irradiation examination of Al-61 wt% U3Si fuel rods from the NRU reactor

International Nuclear Information System (INIS)

Sears, D.F.; Wang, N.

1997-01-01

This paper describes the post-irradiation examination of 4 intact low enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 O coolant inlet temperature 37E C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 : m thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 : m thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on Al-61 wt% U 3 Si fuel irradiated in the NRU reactor. (author)

Post-Irradiation Examination of 237Np Targets for 238Pu Production

Energy Technology Data Exchange (ETDEWEB)

Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Hobbs, Randy W [ORNL; Schmidlin, Joshua E [ORNL

2015-01-01

Oak Ridge National Laboratory is recovering the US ²³⁸Pu production capability and the first step in the process has been to evaluate the performance of a ²³⁷Np target cermet pellet encased in an aluminum clad. The process proceeded in 3 steps; the first step was to irradiate capsules of single pellets composed of NpO₂ and aluminum power to examine their shrinkage and gas release. These pellets were formed by compressing sintered NpO₂ and aluminum powder in a die at high pressure followed by sintering in a vacuum furnace. Three temperatures were chosen for sintering the solution precipitated NpO₂ power used for pellet fabrication. The second step was to irradiate partial targets composed of 8 pellets in a semi-prototypical arrangement at the two best performing sintering temperatures to determine which temperature gave a pellet that performed the best under the actual planned irradiation conditions. The third step was to irradiate ~50 pellets in an actual target configuration at design irradiation conditions to assess pellet shrinkage and gas release, target heat transfer, and dimensional stability. The higher sintering temperature appeared to offer the best performance after one cycle of irradiation by having the least shrinkage, thus keeping the heat transfer gap between the pellets and clad small minimizing the pellet operating temperature. The final result of the testing was a target that can meet the initial production goals, satisfy the reactor safety requirements, and can be fabricated in production quantities. The current focus of the program is to verify that the target can be remotely dissembled, the pellets dissolved, and the ²³⁸Pu recovered. Tests are being conducted to examine these concerns and to compare results to code predictions. Once the performance of the full length targets has been quantified, the pellet ²³⁷Np loading will be revisited to determine if it can be

Design of a decontamination section of the post-irradiation examination laboratory

International Nuclear Information System (INIS)

Homberger, Victor; Coronel, Ruben R.; Laumann, Victor; Perez, Jorge O.; Quinteros, Andrea N.; Ratner, Marcos

1999-01-01

The Post-Irradiation Examination Laboratory activities include the decontamination of expensive equipment of different sizes and weight, involving the complexity and extension of the necessary decontamination. A decontamination section has been designed for that purpose. (author)

Construction of irradiated material examination facility-basic design

International Nuclear Information System (INIS)

Ro, Seung Gy; Kim, Eun Ka; Hong, Gye Won; Herr, Young Hoi; Hong, Kwon Pyo; Lee, Myeong Han; Baik, Sang Youl; Choo, Yong Sun; Baik, Seung Je

1989-02-01

The basic design of the hot cell facility which has the main purpose of doing mechanical and physical property tests of irradiated materials, the examination process, and the annexed facility has been made. Also basic and detall designs for the underground excavation work have been performed. The project management and tasks required for the license application have been carried out in due course. The facility is expected to be completed by the end of 1992, if the budgetary support is sufficient. (Author)

Examination of Experimental Data for Irradiation - Creep in Nuclear Graphite

Science.gov (United States)

Mobasheran, Amir Sassan

The objective of this dissertation was to establish credibility and confidence levels of the observed behavior of nuclear graphite in neutron irradiation environment. Available experimental data associated with the OC-series irradiation -induced creep experiments performed at the Oak Ridge National Laboratory (ORNL) were examined. Pre- and postirradiation measurement data were studied considering "linear" and "nonlinear" creep models. The nonlinear creep model considers the creep coefficient to vary with neutron fluence due to the densification of graphite with neutron irradiation. Within the range of neutron fluence involved (up to 0.53 times 10^{26} neutrons/m ^2, E > 50 KeV), both models were capable of explaining about 96% and 80% of the variation of the irradiation-induced creep strain with neutron fluence at temperatures of 600^circC and 900^circC, respectively. Temperature and reactor power data were analyzed to determine the best estimates for the actual irradiation temperatures. It was determined according to thermocouple readouts that the best estimate values for the irradiation temperatures were well within +/-10 ^circC of the design temperatures of 600^circC and 900 ^circC. The dependence of the secondary creep coefficients (for both linear and nonlinear models) on irradiation temperature was determined assuming that the variation of creep coefficient with temperature, in the temperature range studied, is reasonably linear. It was concluded that the variability in estimate of the creep coefficients is definitely not the results of temperature fluctuations in the experiment. The coefficients for the constitutive equation describing the overall growth of grade H-451 graphite were also studied. It was revealed that the modulus of elasticity and the shear modulus are not affected by creep and that the electrical resistivity is slightly (less than 5%) changed by creep. However, the coefficient of thermal expansion does change with creep. The consistency of

Radiation protection aspects in the metallurgical examination of irradiated fuel elements

International Nuclear Information System (INIS)

Janardhanan, S.; Pillai, P.M.B.; Jacob, John; Kutty, K.N.; Wattamwar, S.B.; Mehta, S.K.

1981-01-01

The operational safety requirements of hot cell facilities for metallurgical examination of irradiated natural and enriched uranium fuel elements are highlighted. The cell shielding is designed for handling activities equivalent of 10 2 to 10 5 curies of gamma energy of 1.3 Mev. A brief outline of the built-in design features relevant to safety assessment is also incorporated. Reference is made to some salient features of Radiometallurgy Cells at Trombay. Metallurgical operations include investigations on cladding failure of irradiated material structure and specimen preparation from hot fuel element. The radiation protection aspects presented in this paper show that handling low irradiated fuel elements in these beta-gamma cells do not cause serious operational safety problems. The procedures followed and the containment provided would adequately restrict exposure of operational staff to acceptable limits. (author)

Detection of rhodopsin dimerization in situ by PIE-FCCS, a time-resolved fluorescence spectroscopy.

Science.gov (United States)

Smith, Adam W

2015-01-01

Rhodopsin self-associates in the plasma membrane. At low concentrations, the interactions are consistent with a monomer-dimer equilibrium (Comar et al., J Am Chem Soc 136(23):8342-8349, 2014). At high concentrations in native tissue, higher-order clusters have been observed (Fotiadis et al., Nature 421:127-128, 2003). The physiological role of rhodopsin dimerization is still being investigated, but it is clear that a quantitative assessment is essential to determining the function of rhodopsin clusters in vision. To quantify rhodopsin interactions, I will outline the theory and methodology of a specialized time-resolved fluorescence spectroscopy for measuring membrane protein-protein interactions called pulsed-interleaved excitation fluorescence cross-correlation spectroscopy (PIE-FCCS). The strength of this technique is its ability to quantify rhodopsin interactions in situ (i.e., a live cell plasma membrane). There are two reasons for restricting the scope to live cell membranes. First, the compositional heterogeneity of the plasma membrane creates a complex milieu with thousands of lipid, protein, and carbohydrate species. This makes it difficult to infer quaternary interactions from detergent solubilized samples or construct a model phospholipid bilayer that recapitulates all of the interactions present in native membranes. Second, organizational structure and dynamics is a key feature of the plasma membrane, and fixation techniques like formaldehyde cross-linking and vitrification will modulate the interactions. PIE-FCCS is based on two-color fluorescence imaging with time-correlated single-photon counting (TCSPC) (Becker et al., Rev Sci Instrum 70:1835-1841, 1999). By time-tagging every detected photon, the data can be analyzed as a fluorescence intensity distribution, fluorescence lifetime histogram, or fluorescence (cross-)correlation spectra (FCS/FCCS) (Becker, Advanced time-correlated single-photon counting techniques, Springer, Berlin, 2005). These

Effect of the enriched filter pie with natural phosphate and microorganisms on soil and plant in a Vermelho-Distrófico Latossolo

Directory of Open Access Journals (Sweden)

Leónides Castellanos González

2016-03-01

Full Text Available The objective of the research was to evaluate the use of filter pie and phosphate rock enriched with biofertilizer microorganisms on the population of microorganisms and phosphorus tenor in the soil and on plants. The experiment was conducted in a vegetation house with corn plants with Latossolo Red-Distrófico, containing seven treatments disposed in an entirely randomized design, where the treatments were: soil samples and filter pie with additions of phosphate of Araxá rock enriched with microorganisms coming from Biopack (1, Embrafos (2 and Azotofos (3, and without enrichment with microorganisms (4, and three controls, soil just with filter pie (5, soil just with phosphate rock (6 and only soil (7, with three repetitions. Height, stem diameter and number of leaves for plants were evaluated at 14, 28 and 42 days. At the 45 days the dry matters of leafs and roots were evaluated, as the same as populations of the microorganisms and phosphorus tenor in the soil and in the aerial part of the plant. The tenor of soluble phosphorus in the soil and accumulated phosphorous in the aerial part of corn plan were increased by the treatments with filter pie but not higher with filter pie enriched with biofertilizers, so its use isn’t justified. The pie treatments increase the fungus, total and solubilizers bacterial populations in the soil, provoking an increase of the height, diameter, number of leaves and dry matter of corn plants in a Red-Distrófico Latossolo.

Técnica de Imhäuser en el pie zambo : resultados a medio y largo plazo

OpenAIRE

Fernández Gutiérrez, C.; Núñez Batalla, Daniel; Pena Vázquez, J.; Antuña Antuña, S.; González Méndez, J.; López Fernández, P; Paz Jiménez, José

1996-01-01

Se revisan un total de 60 pies con equinovaro congénito pertenecientes a 48 pacientes, todos con un seguimiento superior a 5 años. En todos los casos se realizó un tratamiento combinado: ortopédico (manipulaciones y enyesados desde el nacimiento) y quirúrgico (correcciones según el método de Imhäuser modificado). Se valoró morfológica, funcional y radiográficamente cada caso. En la valoración global sólo 1 caso fue considerado malo, 13 regulares (21,6%), 6 por pies planos valgo...

Safety analysis report of the irradiation test of Type-B bundle

Energy Technology Data Exchange (ETDEWEB)

Lee, Choong Sung; Lim, I. C.; Lee, B. C.; Ryu, J. S.; Kim, H. R

2000-06-01

The HANARO fuel, U{sub 3}Si-A1, has been developed by AECL and tested in NRU reactor. In the course of the fuel qualification tests, only one case was performed under the higher power condition than maximum linear power which was expected in the design stage. The Korea regulatory body, KINS imposed that HANARO shall be operated at the power level less than 24MW which is 80% of the design full power until HANARO shows the repetitive performance of the fuel at the power condition abov e 112.8KW/m. To resolve this imposition, KAERI designed two types of special test bundles: two non-instrumented(Type-A) and one instrumented(Type-B) test bundles. Two Type-A bundles were irradiated in HANARO: one of them has finished PIE and the other is under PIE. Type-B bundle was loaded in the core during 1.32 day at 1996, but outstanding FIV(flow induced vibration) was observed at the pool top because of long guide tube attached to the top of the bundle. The successful installation of the chimney fastener to fix the guide tube resulted in conducting the irradiation test of Type-B bundle again. The test will start at mid- July, 2000. In order to safely do the Type-B irradiation test, the safety analysis for the nuclear, mechanical and thermal-hydraulic aspects was performed. The reactivity worth and the maximum 1 near power predicted by VENTURE are 6.3mk/k and 121.6kW/m, respectively. Thermal margins for normal and transient conditions using MATRA-h, are assessed to satisfy the safety criteria.

Radiation protection aspects in the metallurgical examination of irradiated fuel elements

Energy Technology Data Exchange (ETDEWEB)

Janardhanan, S.; Pillai, P.M.B.; Jacob, J.; Kutty, K.N.; Wattamwar, S.B.; Mehta, S.K. (Bhabha Atomic Research Centre, Bombay (India). Health Physics Div.)

The operational safety requirements of hot cell facilities for metallurgical examination of irradiated natural and enriched uranium fuel elements are highlighted. The cell shielding is designed for handling activities equivalent of 10/sup 2/ to 10/sup 5/ curies of gamma energy of 1.3 Mev. A brief outline of the built-in design features relevant to safety assessment is also incorporated. Reference is made to some salient features of Radiometallurgy Cells at Trombay. Metallurgical operations include investigations on cladding failure of irradiated material structure and specimen preparation from hot fuel element. The radiation protection aspects presented in this paper show that handling low irradiated fuel elements in these beta-gamma cells do not cause serious operational safety problems. The procedures followed and the containment provided would adequately restrict exposure of operational staff to acceptable limits.

Postirradiation examination results for the Irradiation Effects Test IE-5

International Nuclear Information System (INIS)

Cook, T.F.; Ploger, S.A.; Hobbins, R.R.

1978-03-01

The results are presented of the postirradiation examination of four pressurized water reactor type fuel rods which were tested in-pile under a fast power ramp and film boiling operation during Irradiation Effects (IE) Test 5. The major objectives of this test were to evaluate the effects of simulated fission products on fuel rod behavior during a fast power ramp, to determine the effects of high initial internal pressure on a fuel rod during film boiling, and to assess fuel rod property changes that occur during film boiling in a fuel rod with previously irradiated cladding. The overall condition of the rods and changes that occurred in fuel and cladding as a result of the power ramp and film boiling operation, as determined from the postirradiation examination, are reported and analyzed. Effects of the simulated fission products on fuel rod behavior during a power ramp are discussed. The effect of high internal pressure on rod behavior during film boiling is evaluated. Cladding temperatures are estimated at various axial and circumferential locations. Cladding embrittlement by oxidation is also assessed

TEM investigation of irradiated U-7 weight percent Mo dispersion fuel

International Nuclear Information System (INIS)

Van den Berghe, S.

2009-01-01

In the FUTURE experiment, fuel plates containing U-7 weight percent Mo atomized powder were irradiated in the BR2 reactor. At a burn-up of approximately 33 percent 235 U (6.5 percent FIMA or 1.41 10 21 fissions/cm 3 meat), the fuel plates showed an important deformation and the irradiation was stopped. The plates were submitted to detailed PIE at the Laboratory for High and Medium level Activity. The results of these examinations were reported in the scientific report of last year and published in open literature. Since then, the microstructural aspects of the FUTURE fuel were studied in more detail using transmission electron microscopy (TEM), in an attempt to understand the nature of the interaction phase and the fission gas behavior in the atomized U(Mo) fuel. The FUTURE experiment is regarded as the definitive proof that the classical atomized U(Mo) dispersion fuel is not stable under irradiation, at least in the conditions required for normal operation of plate-type fuel. The main cause for the instability was identified to be the irradiation behavior of the U(Mo)-Al interaction phase which is formed between the U(Mo) particles and the pure aluminum matrix during irradiation. It is assumed to become amorphous under irradiation and as such cannot retain the fission gas in stable bubbles. As a consequence, gas filled voids are generated between the interaction layer and the matrix, resulting in fuel plate pillowing and failure. The objective of the TEM investigation was the confirmation of this assumption of the amorphisation of the interaction phase. A deeper understanding of the actual nature of this layer and the fission gas behaviour in these fuels in general can allow a more oriented search for a solution to the fuel failures

The second Euratom sponsored 9000C HTR fuel irradiation experiment in the HFR Petten Project E 96.02: Pt.2. Post-irradiation examination

International Nuclear Information System (INIS)

Roettger, R.; Bueger, J. de; Schoots, T.

1977-01-01

A large variety of HTR fuel specimens, loose coated particles, coupons and compacts provided by Belgonucleaire, the Dragon Project and the KFA Juelich have been irradiated in the HFR at Petten at about 900 0 C up to a maximum fast neutron fluence of about 7x10 21 cm -2 (EDN) as a Euratom sponsored experiment. The maximum burn-ups were between 11 and 18.5% FIMA. The results of the post-irradiation examinations, comprising visual inspection, dimensional measurements, microradiography, metallography, and burn-up determinations are presented in this part 2 of the final report. The examinations have shown that the endurance limit of most of the tested fuel varieties is beyond the reached irradiation values

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Post-irradiation examination of A1-61 wt % U3Si fuel rods from the NRU reactor

International Nuclear Information System (INIS)

Sears, D.F.; Wang, N.

1997-09-01

This paper describes the post-irradiation examination of 4 intact low-enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 0 coolant inlet temperature 37 degrees C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 μm thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 μm thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on A1-61 wt % U 3 Si fuel irradiated in the NRU reactor. (author)

Atritis gotosa como causa de pie de riesgo

OpenAIRE

Ortega-Avila, Ana Belén; Fernández Torres, Raúl; Agüera Jiménez, David

2016-01-01

INTRODUCCIÓN: La gota es una manifestación común de la artritis, cuya inflamación se debe al depóstido de cristales de uratao monosódico en articulaciones y tejidos blandos por un aumento de ácido úrico en sangre y su padecimiento muestra una relación creciente con la edad. OBJETIVOS: Analizar las características fisiopatológicas de la artritis gotosa en el pie así como justificar la repercusión negativa que tiene su padecimiento en la calidad de vida de la población basándose en la simple af...

Postirradiation examination results for the Irradiation Effects Test 2

International Nuclear Information System (INIS)

Ploger, S.A.; Kerwin, D.K.; Croucher, D.W.

1978-01-01

This report presents the postirradiation examination results of Test IE-2 in the Irradiation Effects Test Series conducted under the Thermal Fuels Behavior Program. The objectives of this test were to evaluate the influence of previous cladding irradiation and fuel-cladding diametral gap on fuel rod behavior during a power ramp and during film boiling operation. Test IE-2, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed two 0.97-m-long pressurized water reactor type fuel rods fabricated from previously irradiated zircaloy-4 cladding and two similar rods fabricated from unirradiated cladding. The four rods were subjected to a preconditioning period, followed by a power ramp to an average peak rod power of 68 kW/m and steady state operation for one hour at an individual rod coolant mass flux of 4880 kg/s . m 2 . After a flow reduction to 2550 kg/s . m 2 , film boiling occurred on three rods. An additional flow reduction to 2245 kg/s . m 2 produced film boiling on the remaining fuel rod. Maximum time in film boiling was 90 s. None of the four fuel rods failed during the test. Damage caused by film boiling, as characterized by oxidation, oxide spalling, and collapse at fuel pellet interfaces, was found on all four rods. Film boiling regions on these rods showed evidence of fuel melting, fuel centerline void formation, and internal cladding oxidation resulting from fuel-cladding reaction. Effects of fuel-cladding diametral gap and cladding irradiation are summarized. Measured temperatures and metallographically estimated temperatures are compared at several axial fuel rod locations

Stuffed Turkey and Pumpkin Pie: In, Through and Out of American Contexts

Directory of Open Access Journals (Sweden)

Tamara Kohn

2013-02-01

Full Text Available This article explores the meanings embedded in the production, consumption, and symbolic positioning of turkey and pumpkin pie, foods closely associated with the American ritual feast of Thanksgiving. An analysis of turkey and pumpkin pie recipes used and adapted by first- and second-generation immigrants in north America, and by north Americans living abroad, throws into relief complex relations between food production, food consumption and the complexities of lived and often multiple sociocultural identifications. Through sharing the experiences, memories and associations evoked by individuals in the production of holiday recipes, I argue that ideas about ‘tradition’ and a desire to celebrate family and community through the ritual of baking, serving and consuming a ‘standard’ Thanksgiving holiday meal allow one to feel part of an imagined global American community. At the same time these details demonstrate celebration of individual and familial distinctiveness that is traced to (sometimes contested memories of childhood and/or ethnic background, as well as to exploration, innovation and experience in the world at large through travel, migration and imagination.

The behaviour under irradiation of molybdenum matrix for inert matrix fuel containing americium oxide (CerMet concept)

Energy Technology Data Exchange (ETDEWEB)

D' Agata, E., E-mail: elio.dagata@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); Knol, S.; Fedorov, A.V. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands); Fernandez, A.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Klaassen, F. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands)

2015-10-15

Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors or Accelerator Driven System (ADS, subcritical reactors dedicated to transmutation) of long-lived nuclides like {sup 241}Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. In order to safely burn americium in a fast reactor or ADS, it must be incorporated in a matrix that could be metallic (CerMet target) or ceramic (CerCer target). One of the most promising matrix to incorporate Am is molybdenum. In order to address the issues (swelling, stability under irradiation, gas retention and release) of using Mo as matrix to transmute Am, two irradiation experiments have been conducted recently at the High Flux Reactor (HFR) in Petten (The Netherland) namely HELIOS and BODEX. The BODEX experiment is a separate effect test, where the molybdenum behaviour is studied without the presence of fission products using {sup 10}B to “produce” helium, the HELIOS experiment included a more representative fuel target with the presence of Am and fission product. This paper covers the results of Post Irradiation Examination (PIE) of the two irradiation experiments mentioned above where molybdenum behaviour has been deeply investigated as possible matrix to transmute americium (CerMet fuel target). The behaviour of molybdenum looks satisfying at operating temperature but at high temperature (above 1000 °C) more investigation should be performed.

Design, irradiation, and post-irradiation examination of the UC and (U,Pu)C fuel rods of the test groups Mol-11/K1 and Mol-11/K2

International Nuclear Information System (INIS)

Freund, D.; Elbel, H.; Steiner, H.

1976-06-01

The test groups K1 and K2 of the irradiation experiment Mol-11 are reported. Design, irradiation, and post-irradiation examination of the fuel rods irradiated are described. Mol-11/K1 consisted of one fuel rod with UC of 94% T.D. and helium bonding. This test group was intended to prove the high power irradiation capsule in pile. Mol-11/K2 consists of three fuel rods in total. One of these is presently still in the reactor. In this test group mixed carbide fuel of 83% T.D. and 15% Pu content under helium bonding is irradiated. The fuel rod K2-2 was provided with a capillary tube for the continuous measurement of fission gas pressure built up. 1.4988 stainless steel was chosen as cladding material. The final burnup lies between 35 and 70 MWd/kg M. Post-irradiation examination of the two test groups covers a theoretical analysis of the irradiation behaviour. (orig./GSCH) [de

Post-irradiation examination of a 13000C-HTR fuel experiment Project J 96.M3

International Nuclear Information System (INIS)

Bueger, J. de; Roettger, H.

1977-01-01

A large variety of loose coated fuel particles have been irradiated in the BR2 at Mol/Belgium at temperatures between 1200 0 C and 1400 0 C and up to a fast neutron fluence of 1.2x1022 cm -2 (E>0.1 MeV) as a Euratom sponsored experiment for the advanced testing of HTR fuel. The specimens have been provided by Belgonucleaire and the Dragon Project. A short description of the experiment as well as the results of post-irradiation examination mainly carried out at Petten (N.H.), The Netherlands, are presented here. The post-irradiation examination has shown that the required performance can be achieved by a number of the tested fuel specimens without serious damage

The post irradiation examination of three fuel rods from the IFA 429 experiment irradiated in the Halden Reactor

International Nuclear Information System (INIS)

Williams, J.

1979-11-01

A series of fuel rod irradiation experiments were performed in the Halden Heavy Boiling Water Reactor in Norway. These were designed to provide a range of fuel property data as a function of burn-up. One of these experiments was the IFA-429. This was designed to study the absorption of helium filling gas by the UO 2 fuel pellets, steady state and transient fission gas release and fuel thermal behaviour to high burn-up. This data was to be obtained as a function of fuel density, fuel grain size, initial fuel/cladding gap, average linear heat rating, burn-up and overpower transients. All the fuel is in the form of pressed and sintered UO 2 pellets enriched to 13 weight percent 235 U. All the rods were clad in Zircaloy 4 tube. The details of the experiment are given. The post irradiation examination included: visual examination, neutron radiography, dimensional measurements, gamma scanning, measurement of gases in fuel rods and internal free volume, burn-up analysis, metallographic examination, measurement of retained gas in UO 2 pellets, measurement of bulk density of UO 2 . The results are given and discussed. (U.K.)

Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

Energy Technology Data Exchange (ETDEWEB)

Paul A Demkowicz; Paul Demkowicz; David V Laug

2010-10-01

Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

Shielding Studies for Reducing the associated Radiological Risks Due To Irradiated Low Enriched Uranium Foil

International Nuclear Information System (INIS)

Margeanu, C.A.

2011-01-01

Present work estimates the radiation dose rates corresponding to irradiated Low Enriched Uranium (20 wt % 235 U) foil as part of shielding studies for radiological risks reduction after irradiation inside TRIGA 14 MW Research Reactor in an investigation on 99 Mo production possibility. Post-Irradiation Examination Laboratory's cell shielding calculations have been performed; radiation source was obtained by using ORIGEN-S code with specific cross-sections libraries. Different post-irradiation cooling times have been considered, gamma dose rates being estimated by using MAVRIC module from Scale 6 programs package, for following exposure situations (relative to Pie cell): i) front side, ii) lateral side and iii) back side. Three different calculations were performed: a) without any protection shield between operator and cell, except for the cell stainless steel wall; b) with a Lead protection shield between operator and cell and c) with a depleted Uranium shield, located inside the cell in between the radiation source and cell window. Radiation dose rates to cell external wall surface and for other eight fixed distances from cell wall were estimated. To obtain a consistent set of solutions, the study was done for various Uranium foil weights and different Lead and depleted Uranium shields thicknesses. Calculations were focused to assure that the dose rate to an operator positioned at 60 cm working distance from the cell will not exceed 0.02 mSv/h, maximum allowed dose rate for professionally exposed personnel according to Romanian regulations.

Introducing a Framework for Physics Innovation and Entrepreneurship (PIE) Education

Science.gov (United States)

Roughani, Bahram

A desired outcome for Physics Innovation and Entrepreneurship (PIE) education is preparing physics majors with an innovative and entrepreneurial mindset who are capable of opportunity recognition and adept in leveraging physics knowledge to address specific needs. Physics as a discipline is well-recognized to prepare students who become problem solvers and critical thinkers, gifted in dealing with abstract ideas and ambiguities in the context of complex and real-world problems. These characteristics when enhanced through appropriate combinations of curricular, co-curricular, and extra-curricular programs can prepare physics majors for careers and future challenges that may involve translating physics knowledge into useful products and services either as part of a technical team within an organization or through startups. A viable PIE education model prepares graduates for various career paths in addition to the traditional options such as pursuing graduate studies or becoming a science teacher. Having a well-defined ``third option'' for physics will benefit the robustness of the physics discipline through recruitment and retention of prospective students who in principle are interested in physics as a subject, but in practice they may overlook physics as their preferred major primarily because they are uncertain about a viable career path based on an undergraduate physics education. The ''Pathways to Innovation'' at Loyola is established based on the program developed by VentureWell and Epicenter (NSF Supported).

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

Grover, S. Blaine

2009-01-01

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

Miniplates irradiation in the ATR (Idaho, USA)

International Nuclear Information System (INIS)

Pasqualini, Enrique E.

2007-01-01

High density U Mo alloys are promising for its utilization in the reconversion of HEU fuels to LEU for research nuclear reactors. Ought to the thermomechanical properties of the alloy U Mo and its interaction with aluminium it is necessary to develop new technologies and fabrication procedures to qualify this material as a nuclear fuel. In this work a review is made about the evolution of the idea and PIE experiments of monolithic LEU U 7 Mo fuel with Zr-4 cladding. The irradiation took place in the frame of international qualification efforts of dispersed and monolithic U Mo fuels. Dispersed and monolithic fuels, elaborated and in intermediate steps of development, are discussed. (author) [es

Validation of CBZ code system for post-irradiation examination analysis and sensitivity analysis of (n,γ) branching ratio

International Nuclear Information System (INIS)

Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2013-01-01

A code system CBZ is being developed in Hokkaido University. In order to validate it, PIE data, which are nuclide composition data of a spent fuel, have been analyzed with CBZ. The validity is evaluated as ratios of the calculation values to the experimental ones (C/E ratios). Differences between experimental values and calculation ones are smaller than 20% except some nuclides. Thus this code system is validated. Additionally, we evaluate influence of change of (n,γ) branching ratio on inventories of fission products and actinides. As a result, branching ratios of Sb-121, Pm-147, and Am-241 influence inventories of several nuclides. We perform PIE analysis using different (n,γ) branching ratio data from the ORIGEN-2 library, JNDC-Ver.2, and JEFF-3.1A, and find that differences in (n,γ) branching ratios between different nuclear libraries have a non-negligible influence on inventories of several nuclides. (author)

VVER fuel. Results of post irradiation examination

International Nuclear Information System (INIS)

Smirnov, V.P.; Markov, D.V.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

2005-01-01

The present paper presents the main results of post-irradiation examination of more than 40 different fuel assemblies (FA) operated in the cores of VVER-1000 and VVER-440-type power reactors in a wide range of fuel burnup. The condition of fuel assembly components from the viewpoint of deformation, corrosion resistance and mechanical properties is described here. A serviceability of the FA design as a whole and interaction between individual FA components under vibration condition and mechanical load received primary emphasis. The reasons of FA damage fuel element failure in a wide range of fuel burnup are also analyzed. A possibility and ways of fuel burnup increase have been proved experimentally for the case of high-level serviceability maintenance of fuel elements to provide for advanced fuel cycles. (author)

Calling Where It Counts: Subordinate Pied Babblers Target the Audience of Their Vocal Advertisements.

Science.gov (United States)

Humphries, David J; Finch, Fiona M; Bell, Matthew B V; Ridley, Amanda R

2015-01-01

For territorial group-living species, opportunities to reproduce on the natal territory can be limited by a number of factors including the availability of resources within a territory, access to unrelated individuals, and monopolies on reproduction by dominant group members. Individuals looking to reproduce are therefore faced with the options of either waiting for a breeding opportunity to arise in the natal territory, or searching for reproductive opportunities in non-natal groups. In the cooperatively breeding Southern pied babbler, Turdoides bicolor, most individuals who achieve reproductive success do so through taking up dominant breeding positions within non-natal groups. For subordinate pied babblers therefore, searching for breeding opportunities in non-natal groups is of primary importance as this represents the major route to reproductive success. However, prospecting (where individuals leave the group to search for reproductive opportunities within other groups) is costly and individuals rapidly lose weight when not part of a group. Here we demonstrate that subordinate pied babblers adopt an alternative strategy for mate attraction by vocal advertisement from within their natal territories. We show that subordinates focus their calling efforts on the edges of their territory, and specifically near boundaries with neighbouring groups that have potential breeding partners (unrelated individuals of the opposite sex). In contrast to prospecting, calling individuals showed no body mass loss associated with this behaviour, suggesting that calling from within the group may provide a 'cheap' advertisement strategy. Additionally, we show that subordinates use information regarding the composition of neighbouring groups to target the greatest number of potential mating partners.

A satisfação do cliente : estudo do caso do franchising de fast-food King Pie na Cidade de Maputo em Moçambique

OpenAIRE

Ramos, Vânia Raquel Raimundo

2014-01-01

Dissertação de Mestrado em Gestão/MBA apresentada à Universidade Aberta O presente trabalho tem por objectivo central, estudar a satisfação dos clientes da empresa King Pie na Cidade de Maputo. O King Pie é uma cadeia sul-africana de fast-food, que em Moçambique opera em regime de franchising, estando presente no mercado local desde o ano de 2004. Neste momento, o King Pie dispõe quatro (4) estabelecimentos localizados na Cidade de Maputo com dois (2) investidores no país que detêm o direi...

IFPE/CNEA-MOX-RAMP, CNEA Power Ramp Irradiations with (PHWR) MOX Fuels

International Nuclear Information System (INIS)

Marino, Armando Carlos; Turnbull, J.A.

2000-01-01

Description: The irradiation of the first MOX nuclear fuel rods fabricated in Argentina began in 1986. These experiences were made in the HFR-Petten reactor, Holland. The six rods were fabricated in the a Facility (GAID-CNEA-Argentina). The first rod has been used for destructive pre-irradiation characterization in the KFK (Kernforschungszentrum Karlsruhe), Germany. The second one was a pathfinder for calibrating HFR systems in Petten. Two other rods included pellets doped with iodine. The first contained mostly CsI whilst the second contained elemental iodine. The concentration of iodine was intended to simulate a burn-up of 15000 MWd/ton(M). The power histories were defined from calculations performed with the BACO code. A 15 day cycle was assumed with a power history that induced PCMI during power cycling. The last high power period was maintained until stress corrosion cracking (SCC) was induced. Two further un-doped rods were used in a sub-program named BU15. Here a burn-up of 15000 MWd/ton(M) was achieved at a low power followed by a final power ramp for one of the rods. The ramp was similar to that used for the Iodine test. The HFR irradiation was conducted satisfactorily. The objective was to attempt a correspondence in behaviour between the doped rods and BU15 rods. PIE detected the presence of micro-cracks inside the cladding of the iodine doped rods. Ramping of the BU15 rod was interrupted when an increase of coolant activity was detected. After discharge, a visual inspection of the rod showed the presence of a small circular hole in the cladding. Additional PIE showed that the hole was due to a SCC failure

AGR-2 Irradiation Test Final As-Run Report

Energy Technology Data Exchange (ETDEWEB)

Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

2014-08-01

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR-2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO₂ (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. 2. Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO₂ TRISO (tristructural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO₂ particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S.-produced fuel.

AGR-2 Irradiation Test Final As-Run Report

Energy Technology Data Exchange (ETDEWEB)

Collin, Blaise P. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

2014-08-01

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO₂ (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO₂ TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO₂ particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel.

Examination in hot laboratories of irradiated fuels from fast reactors

International Nuclear Information System (INIS)

Clottes, G.; Peray, R.; Ratier, J.L.

1980-05-01

Low irradiation rate examinations were carried out soon after the Rapsodie, Rapsodie Fortissimo and Phenix reactors were started up for the first time in order to check the level of maximum temperatures reached and the radial migration of oxygen and plutonium and to assess the movements of fuels inside the cladding. The other examinations were effected at a high specific burnup in order to defines the limit specific burnup securing the integrity of the fuel pin claddings (distortion, ruptures and possible consequences). The examinations carried out so far on fuel elements coming from Phenix or Rapsodie have allowed good fuel surveillance to be undertaken and the acquisition of a large number of data, thanks to which the fuel characteristics of future reactors of the system have been developed [fr

Inferred residual gaps for IFA-527 rods, compared to PIE measurements

International Nuclear Information System (INIS)

Lanning, D.D.

1983-05-01

The NRC-sponsored assembly IFA-527 contained six xenon-filled rods, each instrumented with fuel thermocouples at each end. Five of the six rods contained stable UO 2 fuel pellets with a fabricated diametral gap size of 230 microns. The rods were unique among instrumented test rods in that their lifetime peak powers were quite low (less than 20 kW/m). The assembly was removed at low exposure (about 2 MWd/kgM) due to pressure seal failures. The prefailure measured fuel temperatures in the five identical rods were quite similar, and indicated an effective fuel relocation of about 30 to 35 percent of the as-fabricated gap. Examination of rod cross sections in PIE, however, reveals virtually no reduction in the as-fabricated gap. In an attempt to reconcile these two observations, it is demonstrated that pellet eccentricity could account for the inferred effective relocation for these xenon-filled rods. That analysis is extended to an estimate of the proper effective relocation value for low-powered He-filled rods

Acciones educativas de prevención del pie diabético Educational actions to prevent diabetic foot complications

Directory of Open Access Journals (Sweden)

Pedro Vejerano García

2009-08-01

Full Text Available La Diabetes Mellitus y sus complicaciones son causas de hospitalización, invalidez y afectación económica. Este trabajo hace referencia a la experiencia camagüeyana en prevención, seguimiento y control del pie diabético. Dicha experiencia demostró que la pobre educación para prevenir las complicaciones del pie diabético, se convierte en un factor de riesgo sociocultural modificable. Una consulta de mayor alcance como la Clínica del Pie Diabético, pone al territorio en condiciones ventajosas para atender la transición epidemiológica de la diabetes y sus complicaciones.Diabetes Mellitus and its complications are causes of hospitalization, disability, and financial problems. This paper deals with the experience in prevention, monitoring, and control of diabetic foot in the province of Camagüey. This experience proved that a poor education in the prevention of diabetic foot complications is a modifiable sociocultural risk factor. As an example of more comprehensive consultation, the Diabetic Foot Clinic has benefitted the territory in the assistance of Diabetes Mellitus and its complications.

Slow improvements of metal exposure, health- and breeding conditions of pied flycatchers (Ficedula hypoleuca) after decreased industrial heavy metal emissions

International Nuclear Information System (INIS)

Berglund, A.M.M.; Nyholm, N.E.I.

2011-01-01

The environment around metal industries, such as smelters, is often highly contaminated due to continuous deposition of metals. We studied nest box breeding populations of pied flycatchers (Ficedula hypoleuca) in a well-studied pollution gradient from a sulfide ore smelter in Northern Sweden, after reduced aerial metal emissions (by 93-99%) from the smelter. The deposition of arsenic, cadmium, copper and zinc (based on moss samples) reflected the reduced emissions fairly well. However, nestling pied flycatchers had similar concentrations of these elements and mercury in tissues (bone, liver and blood) and feces in the 2000s, as in the 1980s, when the emissions were substantially higher. The exposure to high metal concentrations in the close vicinity of the smelter resulted in inhibited ALAD activities, depressed hemoglobin and hematocrit levels and increased mortality of nestlings. Our results indicate that in the highly contaminated environment around the smelter, nestlings reflected the slowly cycling soil pool, rather than the atmospheric deposition, and the concentration in soils plays an important role for the response of pied flycatchers to reduced atmospheric deposition. - Highlights: → Pied flycatchers were studied in a pollution gradient from a sulfide smelter. → Metal emissions from the smelter have decreased substantially. → Nestling birds still had high metal concentrations in tissues. → Health and survival rates of nestlings were negatively affected. → Recovery of birds is not expected in the near future.

Slow improvements of metal exposure, health- and breeding conditions of pied flycatchers (Ficedula hypoleuca) after decreased industrial heavy metal emissions

Energy Technology Data Exchange (ETDEWEB)

Berglund, A.M.M., E-mail: asa.berglund@emg.umu.se; Nyholm, N.E.I.

2011-09-15

The environment around metal industries, such as smelters, is often highly contaminated due to continuous deposition of metals. We studied nest box breeding populations of pied flycatchers (Ficedula hypoleuca) in a well-studied pollution gradient from a sulfide ore smelter in Northern Sweden, after reduced aerial metal emissions (by 93-99%) from the smelter. The deposition of arsenic, cadmium, copper and zinc (based on moss samples) reflected the reduced emissions fairly well. However, nestling pied flycatchers had similar concentrations of these elements and mercury in tissues (bone, liver and blood) and feces in the 2000s, as in the 1980s, when the emissions were substantially higher. The exposure to high metal concentrations in the close vicinity of the smelter resulted in inhibited ALAD activities, depressed hemoglobin and hematocrit levels and increased mortality of nestlings. Our results indicate that in the highly contaminated environment around the smelter, nestlings reflected the slowly cycling soil pool, rather than the atmospheric deposition, and the concentration in soils plays an important role for the response of pied flycatchers to reduced atmospheric deposition. - Highlights: {yields} Pied flycatchers were studied in a pollution gradient from a sulfide smelter. {yields} Metal emissions from the smelter have decreased substantially. {yields} Nestling birds still had high metal concentrations in tissues. {yields} Health and survival rates of nestlings were negatively affected. {yields} Recovery of birds is not expected in the near future.

49 CFR 40.361 - What is the purpose of a public interest exclusion (PIE)?

Science.gov (United States)

2010-10-01

... transportation employers and employees from serious noncompliance with DOT drug and alcohol testing rules, the Department's policy is to ensure that employers conduct business only with responsible service agents. (b... alcohol testing regulations, has shown that it is not currently acting in a responsible manner. (c) A PIE...

AGR-1 Fuel Compact 6-3-2 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Paul demkowicz; jason Harp; Scott Ploger

2012-12-01

Destructive post-irradiation examination was performed on fuel Compact 6-3-2, which was irradiated in the AGR-1 experiment to a final compact average burnup of 11.3% FIMA and a time-average, volume-average temperature of 1070°C. The analysis of this compact was focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of fuel burnup by several methods, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. A single particle with a defective SiC layer was identified during deconsolidation-leach-burn-leach analysis, which is in agreement with previous measurements showing elevated cesium in the Capsule 6 graphite fuel holder associated with this fuel compact. The fraction of the compact europium inventory released from the particles and retained in the matrix was relatively high (approximately 6E-3), indicating release from intact particle coatings. The Ag-110m inventory in individual particles exhibited a very broad distribution, with some particles retaining =80% of the predicted inventory and others retaining less than 25%. The average degree of Ag-110m retention in 60 gamma counted particles was approximately 50%. This elevated silver release is in agreement with analysis of silver on the Capsule 6 components, which indicated an average release of 38% of the Capsule 6 inventory from the fuel compacts. In spite of the relatively high degree of silver release from the particles, virtually none of the Ag-110m released was found in the compact matrix, and presumably migrated out of the compact and was deposited on the irradiation capsule components. Release of all other fission products from the particles appears to be less than a single

Hazard evaluation of The International Fusion Materials Irradiation Facility

Energy Technology Data Exchange (ETDEWEB)

Burgazzi, Luciano [ENEA-Centro Ricerche ' Ezio Clementel' , Advanced Physics Technology Division, Via Martiri di Monte Sole, 4, 40129 Bologna (Italy)]. E-mail: burgazzi@bologna.enea.it

2005-01-15

The International Fusion Materials Irradiation Facility (IFMIF) is aimed to provide an intense neutron source by a high current deuteron linear accelerator and a high-speed lithium flow target, for testing candidate materials for fusion. Liquid lithium is being circulated through a loop and is kept at a temperature above its freezing point. In the frame of the design phase called Key Element technology Phase (KEP), jointly performed by an international team to verify the most important risk factors, safety assessment of the whole plant has been required in order to identify the hazards associated with the plant operation. This paper discusses the safety assessments that were performed and their outcome: Failure Mode and Effect Analysis (FMEA) approach has been adopted in order to accomplish the task. Main conclusions of the study is that, on account of the safety and preventive measures adopted, potential plant related hazards are confined within the IFMIF security boundaries and great care must be exercised to protect workers and site personnel from operating the plant. The analysis has provided as a result a set of Postulated Initiating Events (PIEs), that is off-normal events, that could result in hazardous consequences for the plant, together with the total frequency and the list of component failures which could induce the PIE: this assures the exhaustive list of major initiating events of accident sequences, helpful to the further accident sequence analysis phase. Finally, for each one of the individuated PIEs, the evaluation of the accident evolution, in terms of effects on the plant and relative countermeasures, has allowed to verify that adequate measures are being taken both to prevent the accident occurrence and to cope with the accident consequences, thus assuring the fulfilment of the safety requirements.

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

Energy Technology Data Exchange (ETDEWEB)

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

The Canonical Immediate Early 3 Gene Product pIE611 of Mouse Cytomegalovirus Is Dispensable for Viral Replication but Mediates Transcriptional and Posttranscriptional Regulation of Viral Gene Products.

Science.gov (United States)

Rattay, Stephanie; Trilling, Mirko; Megger, Dominik A; Sitek, Barbara; Meyer, Helmut E; Hengel, Hartmut; Le-Trilling, Vu Thuy Khanh

2015-08-01

Transcription of mouse cytomegalovirus (MCMV) immediate early ie1 and ie3 is controlled by the major immediate early promoter/enhancer (MIEP) and requires differential splicing. Based on complete loss of genome replication of an MCMV mutant carrying a deletion of the ie3-specific exon 5, the multifunctional IE3 protein (611 amino acids; pIE611) is considered essential for viral replication. Our analysis of ie3 transcription resulted in the identification of novel ie3 isoforms derived from alternatively spliced ie3 transcripts. Construction of an IE3-hemagglutinin (IE3-HA) virus by insertion of an in-frame HA epitope sequence allowed detection of the IE3 isoforms in infected cells, verifying that the newly identified transcripts code for proteins. This prompted the construction of an MCMV mutant lacking ie611 but retaining the coding capacity for the newly identified isoforms ie453 and ie310. Using Δie611 MCMV, we demonstrated the dispensability of the canonical ie3 gene product pIE611 for viral replication. To determine the role of pIE611 for viral gene expression during MCMV infection in an unbiased global approach, we used label-free quantitative mass spectrometry to delineate pIE611-dependent changes of the MCMV proteome. Interestingly, further analysis revealed transcriptional as well as posttranscriptional regulation of MCMV gene products by pIE611. Cytomegaloviruses are pathogenic betaherpesviruses persisting in a lifelong latency from which reactivation can occur under conditions of immunosuppression, immunoimmaturity, or inflammation. The switch from latency to reactivation requires expression of immediate early genes. Therefore, understanding of immediate early gene regulation might add insights into viral pathogenesis. The mouse cytomegalovirus (MCMV) immediate early 3 protein (611 amino acids; pIE611) is considered essential for viral replication. The identification of novel protein isoforms derived from alternatively spliced ie3 transcripts prompted

Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

1986-07-01

Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

A meta-analysis on PewDiePie, his swearing and its development

OpenAIRE

Mononen, O. (Ossi)

2018-01-01

Abstract This thesis discusses the swearing present in the videos of the YouTube content creator PewDiePie through meta-analysis of a research article on the subject. The thesis will approach the primary article from the points of view of the fields of language play, language policy, ecological language learning and linguistic superdiversities. The primary article is analysed from the point of view of each field in order t...

The impact of conflict issues on fixed-pie perceptions, problem solving, and integrative outcomes in negotiation

NARCIS (Netherlands)

Harinck, F.; de Dreu, C.K.W.; van Vianen, A.E.M.

2000-01-01

It is argued that a negotiators fixed-pie perception, cooperative motivation, problem-solving behavior, and integrative outcomes are influenced by the content of the negotiationthe conflict issue. Negotiation involves conflicting interests, conflicting ideas about intellective problems, or

IFPE/TRIBULATION R1, Fuel Rod Behaviour at High Burnup

International Nuclear Information System (INIS)

Turnbull, J.A.

2002-01-01

Description: The TRIBULATION (Tests Relative to High Burnup Limitations Arising Normally in LWRs) International Programme started in July 1980 and was organized jointly by BelgoNucleaire and the Nuclear Energy Centre at Mol (CEN/SCK) with the co-sponsorship of 14 participating organizations. The objectives of the programme were twofold. It was primarily a demonstration programme aimed at assessing the fuel rod behaviour at high burn-up, when an earlier transient had occurred in the power plant. The second objective was to investigate the behaviour of different fuel rod designs and manufacturers when subjected to a steady state irradiation history to high burn-up. The first objective was met by irradiating fuel rods under steady state conditions in the BR3 reactor and under transient conditions in BR2. The effect of the transient was determined by comparing data from 4 identical rods tested as follows: i) BR3 irradiation followed by PIE; ii) BR3 irradiation followed by BR2 transient then PIE; iii) BR3 irradiation followed by BR2 transient and re-irradiated in BR3 before PIE; iv) BR3 irradiation and continued BR3 irradiation to maximum burn-up before PIE. The Database contains data from 19 cases using rods fabricated by BelgoNucleaire (BN) (11) and Brown Boveri Reactor GmbH (BBR) (8)

A radio-crystallographic apparatus for the examination of irradiated materials; Appareil de radiocristallographie pour examens de materiaux irradies

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

Description of a radio-crystallographic apparatus designed for the examination of irradiated metals. - The double diffraction method was used: a monochromator is placed in the beam of X-rays diffracted by the sample, thus enabling lead shielding to be interposed between the sample and the counter. In this way a very large proportion of the radiation due to the activity of the sample is eliminated. By this method substances with activities of the order of several tens of curies may be examined, and the relative precision obtained in measurements of parameters with this apparatus is equal to 10{sup -4} in the case of good crystals. (author) [French] Description d'un appareil de radiocristallographie destine a l'examen de metaux irradies. - Nous avons utilise la methode de double diffraction: un monochromateur est place dans le faisceau de rayons X diffracte par l'echantillon, ce qui permet d'interposer une protection de plomb entre l'echantillon et le compteur. On elimine ainsi dans une tres grande proportion le rayonnement du a la radioactivite de l'echantillon. On peut ainsi examiner des substances d'une activite de l'ordre de quelques dizaines de curies et la precision relative obtenue pour les mesures de parametres avec cet appareil est egale a 10{sup -4} dans le cas de bons cristaux. (auteur)

The US nuclear science user facilities - 5276

International Nuclear Information System (INIS)

Kennedy, J.R.

2015-01-01

The primary mission of the NSUF (Nuclear Science User Facilities) is to provide access, at no cost to the researcher, to world-class, state-of-the art capabilities and expertise to advance nuclear science and technology through high impact research. Through the NSUF, nuclear energy researchers can access specialized and often unique and expensive equipment and facilities, as well as the accompanying expertise, including nuclear test reactors, ion beam accelerators, hot cell post-irradiation examination (PIE) equipment, synchrotron beam lines, and advanced radiologically qualified materials science PIE instrumentation. The NSUF can also support the design and fabrication of an irradiation experiment, the transport of that experiment to and from the reactor, the PIE activities, the analysis and interpretation of the data, and final material disposition. A special feature of the NSUF is its Sample Library of irradiated specimens made available to users that reduces investigation time and costs. Enhancing the Sample Library for future applications of advanced instrumentation and new ideas is a key goal of the NSUF. Similar to the effort on building a Sample Library, the NSUF is creating a searchable database of the infrastructure available to DOE-NE (Department Of Energy - Office of Nuclear Energy) supported researchers

Análisis de los factores influyentes en la prevalencia de las úlceras de pie en pacientes en hemodiálisis

Directory of Open Access Journals (Sweden)

María Jesús Lechuga Domínguez

Full Text Available Resumen Introducción Los pacientes en tratamiento renal sustitutivo con terapia de hemodiálisis, tienen elevado riesgo de sufrir úlceras en los pies, debido a la presencia de numerosos factores de riesgo asociados. Objetivos El presente estudio, se plantea como objetivo principal establecer la prevalencia de úlceras en los pies, en los pacientes que realizan tratamiento con hemodiálisis en el Hospital San Rafael de A Coruña. Como objetivo secundario, valorar los factores de riesgo de ulceración del pie. Material y Método Se realiza un estudio descriptivo transversal, en el que incluyen a los 131 pacientes con insuficiencia renal crónica (IRC en tratamiento con hemodiálisis, que cumplen los criterios de inclusión. Se ha marcado como variable principal del estudio la presencia o ausencia de úlcera en el pie. Se valora el estado vascular, con la palpación de pulsos y con registro de sonido doppler de las arterias pedia y tibial posterior, además se realiza el cálculo del índice tobillo brazo (ITB. Se valora el estado neurológico, tras la exploración de la sensibilidad protectora con monofilamento y de la sensibilidad profunda con diapasón graduado. Resultados El 10,7% de los pacientes del estudio tienen úlcera activa en el pie. No se logra palpar pulso en la arteria pedia en un 43,5% de los pacientes y en la arteria tibial posterior en un 60% de la muestra estudiada. El 25% de los pacientes presenta un valor de ITB, por debajo de 0,90 y un 13,5% presenta arterias no compresibles. Presentan alteración de sensibilidad profunda un 48,9% de los pacientes y alteración de la sensibilidad protectora un 44,3%. Conclusión La prevención primaria a través de protocolos estandarizados de exploración del pie en las unidades de hemodiálisis, podrá minimizar la posibilidad de ulceraciones que comprometan la calidad de vida de los pacientes. Sistemáticamente habrá que realizar una valoración vascular, neurológica y de alteraciones

Irradiated fuel examination using the Cerenkov technique

International Nuclear Information System (INIS)

Nicholson, N.; Dowdy, E.J.

1981-03-01

A technique for monitoring irradiated nuclear fuel inventories located in water filled storage ponds has been developed and demonstrated. This technique provides sufficient qualitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors. Measurements have been made on the Cerenkov glow light intensity from irradiated fuel that show the intensity of this light to be proportional to the cooling time. Fieldable instruments used in several tests confirm that such measurements can be made easily and rapidly, without fuel assembly movement or the introduction of apparatus into the storage ponds. The Cerenkov technique and instrumentation have been shown to be of potential use to operators of reactor spent fuel facilities and away from reactor storage facilities, and to the International Atomic Energy Agency inspectors who provide surveillance of the irradiated fuel stored in these facilities

A proposal for laminated pie mechanical construction of a toroidal magnet for the far detector for the MINOS experiment

International Nuclear Information System (INIS)

Fields, T.; Guarino, V.; Petereit, E.; Schoessow, P.; Thompson, K.

1996-01-01

This proposal describes an alternative to the reference design for the construction of the toroidal magnet for the detector for the MINOS experiment. This design proposes to construct the steel planes from several steel sheets and laminate them into the required thickness of four centimeters. The 8 meter planes are constructed by cutting all of the steel plates to the same size, which is pie a pie shaped segment of either 30 or 22.5 degrees each. All of the plates in the construction are identical, which is conducive to rapid production and lower cost. The advantages of the proposed laminated construction over the reference design are listed in this paper

NRX and NRU reactor research facilities and irradiation and examination charges

International Nuclear Information System (INIS)

1960-08-01

This report details the irradiation and examination charges on the NRX and NRU reactors at the Chalk River Nuclear Labs. It describes the NRX and NRU research facilities available to external users. It describes the various experimental holes and loops available for research. It also outlines the method used to calculate the facilities charges and the procedure for applying to use the facilities as well as the billing procedures.

Pie waste - A component of food waste and a renewable substrate for producing ethanol.

Science.gov (United States)

Magyar, Margaret; da Costa Sousa, Leonardo; Jayanthi, Singaram; Balan, Venkatesh

2017-04-01

Sugar-rich food waste is a sustainable feedstock that can be converted into ethanol without an expensive thermochemical pretreatment that is commonly used in first and second generation processes. In this manuscript we have outlined the pie waste conversion to ethanol through a two-step process, namely, enzyme hydrolysis using commercial enzyme products mixtures and microbial fermentation using yeast. Optimized enzyme cocktail was found to be 45% alpha amylase, 45% gamma amylase, and 10% pectinase at 2.5mg enzyme protein/g glucan produced a hydrolysate with high glucose concentration. All three solid loadings (20%, 30%, and 40%) produced sugar-rich hydrolysates and ethanol with little to no enzyme or yeast inhibition. Enzymatic hydrolysis and fermentation process mass balance was carried out using pie waste on a 1000g dry weight basis that produced 329g ethanol at 20% solids loading. This process clearly demonstrate how food waste could be efficiently converted to ethanol that could be used for making biodiesel by reacting with waste cooking oil. Copyright © 2017 Elsevier Ltd. All rights reserved.

Examination of Beryllium Under Intense High Energy Proton Beam at CERN's HiRadMat Facility

CERN Document Server

Ammigan, K.; Hurh, P.; Zwaska, R.; Atherton, A.; Caretta, O.; Davenne,T.; Densham, C.; Fitton, M.; Loveridge, P.; O'Dell, J.; Roberts, S.; Kuksenko, V.; Butcher, M.; Calviani, M.; Guinchard, M.; Losito, R.

2017-01-01

Beryllium is extensively used in various accelerator beam lines and target facilities as material for beam win- dows, and to a lesser extent, as secondary particle produc- tion targets. With increasing beam intensities of future ac- celerator facilities, it is critical to understand the response of beryllium under extreme conditions to avoid compro- mising particle production efficiency by limiting beam pa- rameters. As a result, the planned experiment at CERN’s HiRadMat facility will take advantage of the test facility’s tunable high intensity proton beam to probe and investigate the damage mechanisms of several grades of beryllium. The test matrix will consist of multiple arrays of thin discs of varying thicknesses as well as cylinders, each exposed to increasing beam intensities. Online instrumentations will acquire real time temperature, strain, and vibration data of the cylinders, while Post-Irradiation-Examination (PIE) of the discs will exploit advanced microstructural characteri- zation and imagin...

Examination of Beryllium Under Intense High Energy Proton Beam at CERN's HiRadMat Facility

CERN Document Server

Ammigan, K; Hurh, P; Zwaska, R; Atherton, A; Caretta, O; Davenne, t; Densham, C; Fitton, M; Loveridge, P; O'Dell, J; Roberts, S; Kuksenko, v; Butcher, M; Calviani, M; Guinchard, M; Losito, R

2015-01-01

Beryllium is extensively used in various accelerator beam lines and target facilities as material for beam win- dows, and to a lesser extent, as secondary particle produc- tion targets. With increasing beam intensities of future ac- celerator facilities, it is critical to understand the response of beryllium under extreme conditions to avoid compro- mising particle production efficiency by limiting beam pa- rameters. As a result, the planned experiment at CERN’s HiRadMat facility will take advantage of the test facility’s tunable high intensity proton beam to probe and investigate the damage mechanisms of several grades of beryllium. The test matrix will consist of multiple arrays of thin discs of varying thicknesses as well as cylinders, each exposed to increasing beam intensities. Online instrumentations will acquire real time temperature, strain, and vibration data of the cylinders, while Post-Irradiation-Examination (PIE) of the discs will exploit advanced microstructural characteri- zation and imagin...

Diseño y manufactura de pie Protésico

OpenAIRE

Hernández Martin, John Alexander; Parra Piñeros, Luis Alberto; Pinzón Pinzón, Cesar Agusto; Bejarano Peña, Oscar Heli; Romero Gutiérrez, Jairo Alberto; García Benavides, Pedro Eladio

2017-01-01

El término prótesis hace referencia al dispositivo que sustituye o compensa la pérdida de un miembro perdido, para el caso objeto de estudio, se da en una prótesis de miembro inferior, en el caso de una amputación de miembros inferiores a nivel transtibial, el pie es el elemento ausente más significativo ya que asume la estabilidad mínima del cuerpo del paciente, además de interferir en el balance intrínseco del cuerpo humano. Por lo anteriormente descrito uno de los pilares más importantes d...

Energy use in the bread, cake, pastry, and pie baking industries

Energy Technology Data Exchange (ETDEWEB)

Cleland, A.C.; Earle, M.D.; Serrallach, G.F.

1980-05-01

A survey was carried out to determine energy consumption in the bread, cake, pastry, and pie baking industry. Information was obtained from 11 bread bakeries. Energy use in the bread bakeries over the period 1977/78 was found to average 2.45 MJ/kg of which 0.46 MJ/kg was electricity and the rest oven and boiler fuels. Process flow charts are shown. The purposes for which electricity are used are described. An energy consumption analysis is presented and energy conservation measures are described. (MCW)

''Cutie Pie,'' A portable radiation instrument

International Nuclear Information System (INIS)

Ballou, C.O.

1947-01-01

A portable beta and gamma radiation meter of exceedingly small dimensions and weight has been developed. User acceptance has been more enthusiastic than any previous instrument of its type. The circuit, using one Victoreen V-32 tube, is the simplest electronic circuit possible for radiation work and gives high sensitivity. Stability exceeds anything of comparable sensitivity which has come to our attention. The short term stability is due to a circuit which prevents emission before the cathode reaches operating temperature. Long term stability has been improved by evacuating the tube enclosure and switch. The complete, one unit instrument, weighs four pounds two ounces, and is carried with a pistol grip. Exclusive of chamber and handle, its dimensions are 3'' wide, 6 1/2'' long, and 5'' high. The case is formed of aluminum and is designed to give excellent visibility of the meter. Three ranges of approximately 50, 500 and 5000 mr/hr have been incorporated in the instruments. The instrument has been named ''Cutie Pie'' due to its diminutive size

The investigation of fast reactor fuel pin start up behaviour in the irradiation experiment DUELL II

International Nuclear Information System (INIS)

Freund, D.; Geithoff, D.

1988-04-01

The irradiation experiments DUELL-II within the SNR-300 operational Transient Experimental Program deal with the investigation of fresh mixed oxide fuel behaviour at start-up. The irradiation has been carried out in the HFR Petten in four so-called DUELL capsules with two fuel pin samples each. The fuel pins with a total length of 453 mm contained a fuel column of 150 mm length, consisting of high dense (U,Pu)O 2-x fuel with an initial porosity of 4%, a Pu-content of 20.9%, and an O/Me ratio of 1.96. The fuel pellet diameter was 6.37 mm, the outer diameter of the SS cladding, material No. 1.4970, was 7.6 mm. The irradiation included four phases, consisting of preconditioning at 85% nominal power (corresponds to 550 W/cm), a following increase to full power, and two following full power periods of 1 and 10 days, respectively. Post irradiation examination showed incomplete fuel restructuring in the first capsules with central void diameters of 800 μm in the hot plane, complete restructuring in the last capsule, leading to central voids of approximately 1 mm diameter. The residual gaps between fuel and clad varied between 25 and 44 μm. The clad inner surface did not show any corrosion attack. The analysis of fuel restructuring has been carried out with the computer code SATURN-S showing good agreement with the PIE results. The analysis led to a series of model improvements, especially for crack volume and relocation modelling. (orig./GL) [de

Post irradiation examination on test fuel pins for PWR

International Nuclear Information System (INIS)

Fogaca Filho, N.; Ambrozio Filho, F.

1981-01-01

Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

Fission gas measuring technology

International Nuclear Information System (INIS)

Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok.

1998-02-01

Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs

Fission gas measuring technology

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok

1998-02-01

Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs.

Lifting the Curse of the Roman: Quintus Horatius Flaccus Meets the Pied Piper of Hamelin.

Science.gov (United States)

Blackburn, William

There are many similarities between Robert Browning's "The Pied Piper of Hamelin" and Lewis Carroll's "Alice in Wonderland" that are seldom noted by literary critics. Both works were begun for the amusement of specific children, both employ a strange subterranean journey as a central device, and both are works of nonsense…

Development and application of PIE apparatuses for high-burnup LWR fuels

International Nuclear Information System (INIS)

Harada, Katsuya; Mita, Naoaki; Nishino, Yasuharu; Amano, Hidetoshi

1999-01-01

The Reactor Fuel Examination Facility (RFEF) is developing the following post irradiation examination apparatuses: Ion Microprobe mass analyzer (IMA), Pellet Thermal Capacity measuring apparatus (PTC), Micro Density Measuring apparatus MDM, Shield-type Field Emission Scanning Electron Microscope (FE-SEM). The present paper mainly describes several technical topics of these apparatuses. (author)

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

International Nuclear Information System (INIS)

Suyama, K.; Mochizuki, H.; Okuno, H.; Miyoshi, Y.

2004-01-01

This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models. (authors)

Post-irradiation examination of prototype Al-64 wt% U3Si2 fuel rods from NRU

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D.

1997-01-01

Three prototype fuel rods containing Al-64 wt% U 3 Si 2 (3.15 gU/cm 3 ) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U 3 Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U 3 Si 2 powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37 degrees C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U 3 Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL's research reactors

Fractographic examination of HT-9 and 9Cr-1Mo Charpy specimens irradiated in the AD-2 test

International Nuclear Information System (INIS)

Gelles, D.S.; Hu, W.L.

1983-01-01

Fracture surface topologies have been examined using scanning electron microscopy for 20 selected half sized Charpy impact specimens of HT-9 and Modified 9Cr-1Mo in order to provide improved understanding of fracture toughness degradation as a result of irradiation for Path E alloys. The specimen matrix included unirradiated specimens and specimens irradiated in EBR-II in the AD-2 experiment. Also, hardness measurements have been made on selected irradiated Charpy specimens. The results of examinations indicate that irradiation hardening due to G-phase formation at 390 0 C is responsible for the large shift in ductile-to-brittle transition temperature (DBTT) found in HT-9. Toughness degradation in HT-9 observed following higher temperature irradiations is attributed to precipitation at delta ferrite stringers. Reductions in toughness as a consequence of irradiation in Modified 9Cr-1Mo are attributed to in-reactor precipitation of (V,Nb)C and M 23 C 6 . It is shown that crack propagation rates for ductile and brittle failure modes can be measured, that they differ by over an order of magnitude and that unexpected multiple shifts in fracture mode from ductile to brittle failure can be attributed to the effect of delta ferrite stringers on crack propagation rates

Program for photon shielding calculations. Examination of approximations on irradiation geometries

International Nuclear Information System (INIS)

Isozumi, Yasuhito; Ishizuka, Fumihiko; Miyatake, Hideo; Kato, Takahisa; Tosaki, Mitsuo

2004-01-01

Penetration factors and related numerical data in 'Manual of Practical Shield Calculation of Radiation Facilities (2000)', which correspond to the irradiation geometries of point isotropic source in infinite thick material (PI), point isotropic source in finite thick material (PF) and vertical incident to finite thick material (VF), have been carefully examined. The shield calculation based on the PI geometry is usually performed with effective dose penetration factors of radioisotopes given in the 'manual'. The present work cleary shows that such a calculation may lead to an overestimate more than twice larger, especially for thick shield of concrete and water. Employing the numerical data in the 'manual', we have fabricated a simple computer program for the estimation of penetration factors and effective doses of radioisotopes in the different irradiation geometries, i.e., PI, PF and VF. The program is also available to calculate the effective dose from a set of radioisotopes in the different positions, which is necessary for the γ-ray shielding of radioisotope facilities. (author)

Enfermedad de mano, pie y boca en un hospital del Callao, 2016

Directory of Open Access Journals (Sweden)

Milton José Max Rodríguez-Zúñiga

Full Text Available La enfermedad de mano, pie y boca (EMPB es una enfermedad exantemática viral, causada principalmente por Coxsackie A16, con una sintomatología típica consistente en fiebre; exantema pápulo-vesicular en manos, pies y genitales; y un enantema ulceroso en boca. En el verano del 2016 se presentó una diversidad de casos en un hospital del Callao en niños y adultos, con una sintomatología compatible con EMPB; el diagnóstico fue clínico, se aplicó terapia de soporte con resolución final de síntomas. En la última década se han presentado reportes en algunos países con una afectación atípica causada por el Coxsackie A6, produciendo lesiones más extensas y en adultos. Sin embargo, el diagnóstico sigue siendo clínico, solo necesitando confirmación virológica en casos atípicos o cuando el diagnóstico no es claro. La importancia de este reporte radica en describir los casos del Callao ocurridos en el verano del 2016, para servir de apoyo a los profesionales de la salud en el diagnóstico y manejo de pacientes con similar sintomatología.

The relationship of family characteristics and bipolar disorder using causal-pie models.

Science.gov (United States)

Chen, Y-C; Kao, C-F; Lu, M-K; Yang, Y-K; Liao, S-C; Jang, F-L; Chen, W J; Lu, R-B; Kuo, P-H

2014-01-01

Many family characteristics were reported to increase the risk of bipolar disorder (BPD). The development of BPD may be mediated through different pathways, involving diverse risk factor profiles. We evaluated the associations of family characteristics to build influential causal-pie models to estimate their contributions on the risk of developing BPD at the population level. We recruited 329 clinically diagnosed BPD patients and 202 healthy controls to collect information in parental psychopathology, parent-child relationship, and conflict within family. Other than logistic regression models, we applied causal-pie models to identify pathways involved with different family factors for BPD. The risk of BPD was significantly increased with parental depression, neurosis, anxiety, paternal substance use problems, and poor relationship with parents. Having a depressed mother further predicted early onset of BPD. Additionally, a greater risk for BPD was observed with higher numbers of paternal/maternal psychopathologies. Three significant risk profiles were identified for BPD, including paternal substance use problems (73.0%), maternal depression (17.6%), and through poor relationship with parents and conflict within the family (6.3%). Our findings demonstrate that different aspects of family characteristics elicit negative impacts on bipolar illness, which can be utilized to target specific factors to design and employ efficient intervention programs. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

Annual report of JMTR, FY2006. April 1, 2006 - March 31, 2007

International Nuclear Information System (INIS)

2008-03-01

During the FY2006 (April 2006 to March 2007), the Japan Materials Testing Reactor (JMTR) was operated for three operation cycles from 162nd cycle to 165th cycle. Various irradiation tests and post-irradiation examinations (PIEs) were performed for studies on Irradiation Assisted Stress Corrosion Cracking (IASCC) of light-water-reactor internals, development of the fusion blanket, basic materials researches, radioisotope production, and so on. The operation was stopped by 165th cycle according to schedule. And maintenance work and preservation work of the facilities has been started for the refurbishment of the JMTR. Renewal of equipments and aging management of equipments and reactor facilities were carried out according to long-term maintenance plan which was based on periodical evaluation related to maintenance and safety management of reactor facilities. Regarding development on irradiation techniques, the in-situ irradiation tests using load control unit which was developed for in-pile SCC tests of IASCC studies were finished. At the hot laboratory, PIEs for the Radiation Induced Surface Activation (RISA) capsule were performed with the X-ray diffract meter which was moved from the Nuclear Science Research Institute. Under the Arrangement for the Implementation of Cooperative Research Program between the JAEA and the Korea Atomic Energy Research Institute (KAERI), mutual exchange was conducted for information exchange of irradiation and PIE techniques. As for the refurbishment and restart of JMTR, contents of the midterm targets were changed according to unofficial notification of budget in FY2007 by the Ministry of Finance. And it was officially decided to start refurbishment work at FY2007 for restart of the JMTR in FY2011. (author)

Mechanical properties and TEM examination of RAFM steels irradiated up to 70 dpa in BOR-60

Energy Technology Data Exchange (ETDEWEB)

Gaganidze, E., E-mail: Ermile.Gaganidze@kit.edu [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiss, O.J.; Aktaa, J. [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V. [Joint Stock Company ' State Scientific Centre Research Institute of Atomic Reactors' , 433510 Dimitrovgrad-10, Ulyanovsk Region (Russian Federation)

2011-10-01

Mechanical properties of Reduced Activation Ferritic/Martensitic (RAFM) steels were studied after irradiation in BOR-60 reactor to a neutron displacement damage of 70 dpa at 330-340 deg. C. Yield stress and Ductile-to-Brittle-Transition-Temperature of EUROFER97 indicate saturation of hardening and embrittlement. The phenomenological models for description of microstructure evolution and resulting irradiation hardening and embrittlement are discussed. The evolution of yield stress with dose is qualitatively understood within a Whapham and Makin model. Dislocation loops examined in TEM are considered a main source for low-temperature irradiation hardening. The analysis of the fatigue data in terms of the inelastic strain reveals comparable fatigue behaviour both for unirradiated and irradiated conditions, which can be described by a common Manson-Coffin relation. The study of helium effects in B-doped model steels indicated progressive material embrittlement with helium content. Post-irradiation annealing of RAFM steels yielded substantial recovery of mechanical properties.

Local survival of pied flycatcher males and females in a pollution gradient of a Cu smelter

International Nuclear Information System (INIS)

Eeva, T.; Hakkarainen, H.; Belskii, E.

2009-01-01

Survival is one of the most central population measures when the effects of the pollution are studied in natural bird populations. However, only few studies have actually measured rigorous survival estimates on adult birds. In recent years there has been a methodological advance in survival analyses by mark-recapture models. We modelled local survival (including mortality and emigration) with the program MARK in a population of a small insectivorous passerine bird, the pied flycatcher (Ficedula hypoleuca), around a point source of heavy metals. The local survival of females in the polluted area was about 50% lower than in the other areas. Males, however, survived relatively well in the heavily polluted area, but showed somewhat lower survival in the moderately polluted area. Different pollution effects between two sexes might be due to pollution-related differences in reproductive effort in females and males, and/or more intensive uptake of heavy metals by laying females. - Female pied flycatchers (Ficedula hypoleuca) show decreased local survival around a copper smelter.

Post-irradiation examination of Oconee 1 fuel - cycle 1 destructive test phase

International Nuclear Information System (INIS)

1979-07-01

Standard B and W Mark-B (15 x 15) pressurized water reactor fuel rods were destructively examined after one cycle of irradiation in the Oconee 1 reactor. Fuel rod average burnup ranged from 10,603 to 11,270 MWd/mtU for the rods examined. Data obtained included fuel rod extraction loads, rod dimensional changes, cladding tensile properties, fuel pellet gap length, fission product distribution, fission gas and crud composition, fuel densification, chemical burnup analysis, and fuel and cladding microstructure. As expected, parametric changes were well within the design envelope. Superficial corrosion and wear were found at spacer grid contact points. However, the 19 rods examined were structurally sound and exhibited no indications of cladding defects associated with pelletcladding interactions

Ontogenetic variation in the plumage colour of female European Pied Flycatchers Ficedula hypoleuca

OpenAIRE

Potti, Jaime; Canal, David; Camacho, Carlos

2014-01-01

Although variation in the dorsal plumage colour of male European Pied Flycatchers Ficedula hypoleuca has received a great deal of attention, females of the species have been usually considered to be nearly uniformly monochromatic brown. Using reflectance spectrophotometry, we explore the age-dependent variation of plumage colour in females. We disentangle the within- and between-individual effects of this pattern and show a within-individual darkening of the mantle colour with age, whereas di...

Post irradiation test report of irradiated DUPIC simulated fuel

International Nuclear Information System (INIS)

Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

2001-12-01

The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

Associations of leisure-time physical activity with balance and lower extremity strength: a validation of the neuromuscular part of the Physical Activity Pie.

Science.gov (United States)

Lindström, Paula J; Suni, Jaana H; Nygård, Clas-Håkan

2009-07-01

The importance of neuromuscular-type exercise (NME) has been recognized in recent recommendations for public health. However, the knowledge on associations and dose response of different types of leisure-time physical activity (LTPA) with musculoskeletal fitness and health is incomplete. This study evaluated the validity of the NME recommendation for public health introduced by the Physical Activity Pie. Engagement in LTPA and health-related fitness were assessed in 2 consecutive studies with the same adult population age 30 to 69 years (n = 575). Cross-sectional associations between different LTPA types and motor and musculoskeletal fitness were examined by logistic-regression models. Engagement in NME was associated with good static and dynamic balance and lower extremity strength. The highest odds ratios (OR) were found between brisk NME and static balance (most vs least fit OR = 2.39, moderate vs least fit OR = 1.94) and brisk NME and leg strength (more vs least fit OR = 2.10). Some associations were also found between brisk aerobic exercise and good balance. This cross-sectional study suggests that the recommendation for NME in the Physical Activity Pie is valid in terms of balance and leg strength, the 2 major fitness factors related to mobility functioning, especially among aging adults.

Risks of fetus irradiation in X-ray and radioisotope examinations in early stages of pregnancy

International Nuclear Information System (INIS)

Husak, V.; Wiedermann, M.; Susila, M.; Ryznar, V.

1981-01-01

Radiation burden is discussed of the fetus in pregnancy. During organogenesis the fetus is not capable of repairs; irradiation causes growth inhibition and malformations. In the last trimestre, irradiation causes disturbances in the somatic growth, leukemia is found in greater incidence in newly-borns. Doses are tabulated from commonest X-ray examinations and the applications of radiopharmaceuticals. The said doses are only valid for the first five weeks of pregnancy; when the thyroid starts functioning, the total dose retained in the thyroid after the application of radiopharmaceuticals may be higher. (H.S.)

Relating the effects of protein type and content in increased-protein cheese pies to consumers' perception of satiating capacity.

Science.gov (United States)

Marcano, J; Varela, P; Fiszman, S

2015-02-01

Since proteins have been shown to have the highest satiation-inducing effects of all the macronutrients, increasing the protein level is one of the main strategies for designing foods with enhanced satiating capacity. However, few studies analyze the effect that protein addition has on the texture and flavor characteristics of the target food item to relate it to the expected satiating capacity it elicits. The present work studied cheese pies with three levels of soy and whey proteins. Since the protein level altered the rheological behavior of the batters before baking and the texture of the baked pies, the feasibility of adding several protein levels for obtaining a range of final products was investigated. A check-all-that-apply questionnaire containing 32 sensory and non-sensory characteristics of the samples was given to consumers (n = 131) who also scored the perceived samples' satiating capacity. The results showed that the type and content of protein contributed distinctive sensory characteristics to the samples that could be related to their satiating capacity perception. Harder and drier samples (high protein levels) were perceived as more satiating with less perceptible sweet and milky cheese pie characteristic flavors. Soy contributed an off-flavour. These results will contribute to a better understanding of the interrelation of all these factors, aiding the development of highly palatable solid foods with enhanced satiating capacities.

Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-08-15

Austenitic stainless steels are widely used as structural materials alloy in reactor pressure vessel internal components because of their high strength, ductility and fracture toughness. However, exposure due to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term operation in high temperature water environments in nuclear power plants is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) for Type 304 SS in BWR environment. JNES had been conducting IASCC project during from JFY 2000 to JFY 2008, and prepared an engineering database on IASCC. However, the data of crack growth rate (CGR) below the critical fluence level are not sufficient. Therefore, evaluation of neutron irradiation effect project (ENI) was initiated to obtain the CGR data below the critical fluence level, and prepare the SCC growth rate diagram for life time evaluation of core shroud. Test specimens have been irradiated in the OECD/Halden reactor, and the post irradiation experiments (PIE) have been conducting during from JFY 2011 to JFY 2013, finally the modified IASCC guide will be prepared in JFY 2013. (author)

Comparative calculations and operation-to-PIE data juxtaposition of the Zaporozhye NPP, WWER-1000 FA-E0325 fuel rods after 4 years of operation up to ∼49 MWd/kgU burnup

International Nuclear Information System (INIS)

Passage, G.; Stefanova, S.; Scheglov, A.; Proselkov, V.

2006-01-01

Operational and PIE data for the Zaporozhe NPP, FA-E0325, WWER-1000 fuel rods were provided in the OECD NEA IFPE Database and were used to perform comparative calculations among several fuel performance codes. The fuel rods had been irradiated for 4 years of operation up to ∼49 MWd/kg U burnup. The fuel rod operation histories are developed for the PINw99, TRANSURANUS (V1M1J03) and TOPRA-2 codes. The initial state fuel rod parameters are analysed and calculations are carried out. The PIE data enable the comparison of experimental measurement with code-calculated values for cladding elongation (49 rods), FGR and gas pressure (35 rods). Cladding diameter creep-down and gap closure results are juxtaposed as well. The capability of the applied codes correctly to predict the WWER fuel rod performance is shown. The WWER-1000 fuel rod data include initial geometrical and design parameters of the fuel rods, as well as description of the operation regime, NPP unit loading history and PIE results at normal conditions. The data are sufficient for modelling all 312 fuel rod and for comparison of calculations with experimental results for a limited number of fuel rods. The comparison between the calculated and measured results discussed in this paper shows that the codes PINw99, TRANSURANUS and TOPRA-2, are capable of adequate predicting the thermophysical and the mechanical performance of the WWER-1000 fuel rods. The PINw99 code predicts conservative BOL FGR values and conservative gas pressure values in the region of burnups higher than 30 MWd/kg U, which can be explained by the underprediction of the cladding gas inner volume and cladding elongation. The improved version PIN2K (not applied in the present study) predicts much better FGR and gas pressure, though, it is still under development in the high burnup FGR modelling part. In the TRANSURANUS code, there are also areas, where refinements are clearly indicated. They are subjects of the ongoing research projects and

Fuel fabrication and post-irradiation examination

Energy Technology Data Exchange (ETDEWEB)

Venter, P J; Aspeling, J C [Atomic Energy Corporation of South Africa Ltd., Pretoria (South Africa)

1990-06-01

This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

Fuel fabrication and post-irradiation examination

International Nuclear Information System (INIS)

Venter, P.J.; Aspeling, J.C.

1990-01-01

This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

Oxidative stability examination of irradiated swine for samples

International Nuclear Information System (INIS)

Formanek, Z.

1995-01-01

DSC (Differential Scanning Calorimetry), OBM (Oxygen Bomb Method) methods and POV (Peroxide Value) determination was applied to determine the oxidative stability of lard samples irradiated by a Co 60 gamma source at the dose of 0; 0.75; 2.5kGy. Both of the methods seem to be suitable to determine oxidative stability of fats, because in every case lowering oxidative stability was received with increasing irradiation dose. (R.P.). 8 refs., 7 figs., 1 tab

Stovetop Earth Pecan Pie

Science.gov (United States)

Robin, C. M.

2005-12-01

Many fluid mechanical experiments with direct applications to Earth Science are performed with sugary syrups using conceptually straightforward procedures. Corn syrup has indeed proven to be a godsend for those studying convection and related non-linear phenomena. In addition, however, it gives experimentalists a deep physical intuition for the interior workings of hot planets. The basic concepts behind plate tectonics and mantle convection are not difficult; indeed, although they may not be aware of it, most students probably have a basic intuitive understanding of fluid mechanics gained in their daily life. However, the large size and long time scale of geophysical processes may be quite intimidating to young students. Even a simple geophysical experiment requires a complicated array of coolers, heaters and measuring and recording equipment. It is of interest to introduce students to the geodynamical concepts that can be visualized in a high-tech lab using familiar processes and equipment. Using a homemade apparatus and grocery store supplies, I propose using a 'Stove-top Earth pecan pie' to introduce simple geodynamic concepts to middle- and high-school students. The initially cold syrup heats up and the pecans begin to float (continent formation), the syrup begins to convect (mantle convection), and convection slows down after the heat is removed (secular cooling). Even Wilson cycles can be simulated by moving the pan to one side or the other of the stovetop or heating element. The activity formally introduces students to convection and its application to the earth, and makes them think about plate motion, heat transfer, scaling, and experimental procedures. As an added bonus, they can eat their experiments after recess!

Desempenho de cordeiros alimentados com inclusão de torta de macaúba na dieta Performance of lambs fed with macauba pie inclusion in the diet

Directory of Open Access Journals (Sweden)

Rafael Alves de Azevedo

2012-11-01

Full Text Available O objetivo deste trabalho foi avaliar o consumo de nutrientes e o desempenho produtivo e econômico de ovinos confinados da raça Santa Inês, alimentados com dietas com diferentes taxas de inclusão de torta de macaúba. Os animais (24 foram distribuídos em delineamento de blocos ao acaso, nos tratamentos com 0, 100, 200 e 300 g kg-1 de torta de macaúba na matéria seca da dieta, em seis repetições, durante 60 dias de confinamento. A adição de torta de macaúba a taxas crescentes à dieta aumentou linearmente o consumo de proteína bruta, extrato etéreo, fibra em detergente neutro e a conversão alimentar. Não houve diferença no consumo de matéria seca, ganho médio diário e ganho de peso vivo total com a adição da torta. O tratamento com inclusão de 100 g kg-1 apresentou a melhor relação custo:benefício na avaliação das receitas bruta e líquida obtidas com a venda dos animais vivos e abatidos. A adição da torta de macaúba em até 300 g kg-1 da dieta não altera o desempenho produtivo dos animais, entretanto, a inclusão de 100 g kg-1 proporciona melhor retorno econômico.The objective of this work was to evaluate the nutrient uptake and the productive and economic performance of confined Santa Inês lambs fed diets with different levels of macauba pie inclusion . The animals (24 were distributed in a randomized block design, in treatments with 0, 100, 200 and 300 g kg-1 macauba pie in the diet dry matter, with six replicates during 60 days of confinement. The addition of macauba pie at increasing rates to the diet linearly increased the intake of crude protein, ether extract, neutral detergent fiber and feed conversion. There was no difference in dry matter intake, average daily gain and total live weight gain with the macauba pie addition. Diet with 100 g kg-1 macauba pie inclusion showed the best cost:benefit ratio in the evaluation of gross and net revenues from the sale of live and slaughtered animals. The addition

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

Captive breeding and reintroduction of the oriental pied hornbill (Anthracoceros albirostris) in Khao Kheow Open Zoo, Thailand.

Science.gov (United States)

Chaiyarat, Rattanawat; Kongprom, Urarikha; Manathamkamon, Darika; Wanpradab, Sunan; Sangarang, Satean

2012-01-01

This article discusses two related issues of the captive breeding and reintroduction of the Oriental pied hornbill (Anthracoceros albirostris) in order to increase its population in the natural habitats. Oriental pied hornbills were bred in Khao Kheow Open Zoo. Three pairs were separated and kept in breeding cages. Females occupied artificial nests between February and April 2005-2007. Eggs were laid and incubated between February and March each year from 2005 to 2007. Nestlings hatched in late March and left the nest in late April 2005-2007. Each breeding pair was fed with approximately 400 g of food each day. All three pairs reproduced resulting in mature offspring of seven in 2005, six in 2006, and five in 2007. Four of sixteen 3-year-old birds were randomly selected and equipped with a GPS receiver on their backs. Activities of the birds attached and unattached with GPS were not significantly different. The first two birds (one female, one male with GPS) were reintroduced on August 5, 2006, and another nine birds (four birds with GPS) were reintroduced on December 26, 2006. The average home range of these reintroduced birds was 0.13 km(2) . Their foods consisted of wild plants and animals in the home range. The first reintroduced pair was able to breed naturally by laying and hatching eggs in an artificial nest. Two juveniles left the nest in April 2008. These results indicate that both captive breeding and reintroduction are potentially important ways to increase the population of the Oriental pied hornbill in natural habitats. © 2011 Wiley Periodicals, Inc.

Creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composite. Interim report on irradiation examinations: 03M-47AS

International Nuclear Information System (INIS)

Baba, Shin-ichi; Sawa, Kazuhiro; Yamaji, Masatoshi; Matsui, Yoshinori; Ishihara, Masahiro

2007-03-01

The creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composites (c/c composite) are carried out in the Japan Material Testing Reactor (JMTR). This report described 4-items; first item is design/evaluation of the capsule for the irradiation test, second is before irradiation measurements for the residual strain due to manufactured cylindrical c/c composite, and third is also before irradiation measurements of the distance between 2-holes of predrilled in the specimen and last item is examination of analysis for the irradiation creep with thermal gradient by VIENUS Code. The normal creep test is static mechanical load on the specimen in thermal condition, but this creep test under irradiation capsule is thermal stress due to thermal gradient at inside specimen in the thermal condition. Consequently, it is necessary as large as possible thermal gradient in the narrow space of the capsule inside volume. In which the tungsten rod (W-rod) was inserted to the cylindrical c/c composite specimen, for γ-ray heat generation density occurred highly and so maximize the difference temperatures of surface wall between inside and outside wall of the specimen. The measurement method of the deflection due to the irradiation creep of cylindrical c/c composite was adopted as way of ruptured the specimen among the predrilled distance of 2-holes before/after irradiation. Accordingly as the laser dimensional apparatus used to measure the distance between the 2-holes of specimen exactly, easy and untouchable. And also before irradiation measurement of the residual stress due to the manufactured process was estimated by neutron diffraction used Residual Stress Analyzer (RESA) at JRR-3M in JAEA. The irradiation test was finished as total irradiation time, average temperature and neutron dose showed 4189 hours, 873 K and 8.2x10 24 (E>1.0MeV:m -2 ) respectively. The thermal stress was estimated by the difference temperatures of 4

Effect of UV irradiation on Echinaceae purpureae interactions with free radicals examined by an X-band (9.3 GHz) EPR spectroscopy.

Science.gov (United States)

Ramos, Paweł; Pilawa, Barbara

The effect of UVA (315-400 nm) irradiation on Echinaceae purpureae interactions with free radicals was examined by the use of electron paramagnetic resonance (EPR) spectroscopy. The changes of antioxidant properties of E. purpureae with time of UV irradiation from 10 to 110 min (10 min steps) were determined. DPPH as the paramagnetic reference was used in this study. Changes of EPR signals of the reference after interactions with nonirradiated and UV-irradiated E. purpureae were detected. Interactions of the tested E. purpureae samples caused decrease of the EPR signal of DPPH as the result of its antioxidant properties. The decrease of the amplitude of EPR line of DPPH was lower for interactions with UV-irradiated E. purpureae . EPR examination confirmed antioxidant properties of E. purpureae . The weaker antioxidant properties of E. purpureae after UV irradiation were pointed out. E. purpureae should be storage in the dark. The tests bring to light usefulness of electron paramagnetic resonance with microwave frequency of 9.3 GHz (an X-band) in examination of storage conditions of pharmacological herbs.

Nondestructive examination of Oconee 1 fuel assemblies after four cycles of irradiation

International Nuclear Information System (INIS)

Pyecha, T.D.; Mayer, J.T.; Guthrie, B.A. III; Riordan, J.E.

1980-12-01

Five B and W Mark B (15 x 15) pressurized water reactor fuel assemblies were nondestructively examined after four cycles of irradiation in the Oconee 1 reactor. Four of the five assemblies examined had a burnup of 40,000 MWd/mtU; the fifth assembly had a burnup of 36,800 MWd/mtU. This effort is part of a Department of Energy program to improve uranium utilization by extending the burnup of light water reactor fuel. The examinations were conducted in the Oconee 1 and 2 spent fuel storage pool. Data obtained included fuel assembly and fuel rod dimensions, water channel spacings, spacer grid and holddown spring forces, fuel column stack and axial gap lengths, and crud samples. The results indicate that the assemblies performed well through four cycles of operation; all of the data were within design limits

Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-09-15

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U₃Si₂) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U₃Si₂ as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U₃Si₂ fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U₃Si₂. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

New possibilities of the isotope distribution examination in irradiated absorbing materials using secondary ion mass spectrometry method

International Nuclear Information System (INIS)

Goncharenko, Y. D.; Evseev, L.A.; Risovany, V.D.

2005-01-01

The SIMS technique (with using a linear analysis and 2D surface imaging) has been to measure the radial distribution of the boron isotope ratio in the boron carbide pellets irradiated in the fast reactor. It was revealed that a radial distribution of isotope ratio in the boron carbide pellets is significantly different after irradiation in fast and thermal reactors. It was showed the advisability of using ion images for such examinations. (Author)

Trade-offs between sexual advertisement and immune function in the pied flycatcher (Ficedula hypoleuca).

OpenAIRE

Kilpimaa, Janne; Alatalo, Rauno V.; Siitari, Heli

2004-01-01

Good genes models of sexual selection assume that sexual advertisement is costly and thus the level of advertisement honestly reveals heritable viability. Recently it has been suggested that an important cost of sexual advertisement might be impairment of the functioning of the immune system. In this field experiment we investigated the possible trade-offs between immune function and sexual advertisement by manipulating both mating effort and activity of immune defence in male pied flycatcher...

Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

International Nuclear Information System (INIS)

Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

1988-11-01

Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

SSPA Equipment Engineering Feasibility Report

Energy Technology Data Exchange (ETDEWEB)

N.E. Woolstenhulme; C.R. Clark

2011-09-01

In response to a demanding reactor conversion schedule, construction of the Shielded Sample Preparation Area (SSPA) was initiated in 2010 to augment the existing capabilities of the Hot Fuel Examination Facility (HFEF). While HFEF is and will remain the workhorse for post irradiation sample preparation, there is currently a large backlog of Post-Irradiation Examination (PIE) experiments caused by numerous competing projects (this backlog is expected to continue for the foreseeable future). HFEF, in its present configuration also lacks the ability to prepare samples suitable for several of the tests that have been identified for the successful conclusion of the RERTR program; these samples require fine detail machining of irradiated fuel plates.

Current status of JAERI Tokai hot cell facilities

International Nuclear Information System (INIS)

Itami, Hiroharu; Morozumi, Minoru; Yamahara, Takeshi

1992-01-01

JAERI has 4 hot cell facilities in order to examine high radioactive materials. Three of them, the Research Hot Laboratory, the Reactor Fuel Examination Facility and the Waste Safety Testing Facility are located in the JAERI Tokai site, and the rest is the JMTR Hot Laboratory in the Oarai site. The Research Hot Laboratory (RHL) was constructed for post-irradiation examination (PIE), especially nuclear related basic research experiment, such as metallurgical, chemical and mechanical examination on fuels and materials irradiated in research and test reactors. This facility has 10 large dimension concrete and 38 lead cells. At present the RHL is used for various kinds of examinations of high radioactive samples such as fuels of research and test reactors, power reactors and high temperature testing reactor (HTTR), and structural materials. The Reactor Fuel Examination Facility (RFEF) was designed and constructed for carrying out PIE of irradiated full-size fuel assemblies of light water reactors (LWRs). This facility has a storage pool, 8 concrete and 5 lead cells. They are currently used for safety evaluation on high burnup and advanced lWR fuels as part of the national program. The Waste Safety Testing Facility (WASTEF) was designed and constructed for safety research on long-term storage and disposal of high level radioactive wastes, generated by fuel reprocessing. The WASTEF has 5 concrete cells and 1 lead cell. Examinations on the behavior of various long-lived fission products in a glass form and in a canister and, releasing behavior of them out of a canister are carrying out under the condition at storage. (author)

Examination of the safety of gamma-irradiated soya flour

International Nuclear Information System (INIS)

Tencheva, S.; Katsareva, Ts.

1988-01-01

Fat-free soya flour was used as model food for demonstrating the safety of food additives decontaminated with ionizing radiation. The food was irradiated with 10 kGy and included in the daily food of line C 57 BL mice in amount of 35%. Important medical-biological tests were performed for evaluating the potential toxicity of irradiated food products. Results are reported on a study of three generations of mice, which had persistently consumed 35% irradiated soya flour. The parameters studied were: body mass, mass of internal organs, number of newborn animals and their body mass after weaning, reproductive capacity in F 1a and hematological parameters. The results were summarized in five tables

Computational analysis of modern HTGR fuel performance and fission product release during the HFR-EU1 irradiation experiment

Energy Technology Data Exchange (ETDEWEB)

Verfondern, Karl, E-mail: k.verfondern@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Xhonneux, André, E-mail: xhonneux@lrst.rwth-aachen.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Nabielek, Heinz, E-mail: heinznabielek@me.com [Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Allelein, Hans-Josef, E-mail: h.j.allelein@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, Chair for Reactor Safety and Reactor Technology, 52072 Aachen (Germany)

2014-07-01

Highlights: • HFR-EU1 irradiation test demonstrates high quality of HTGR spherical fuel elements. • Irradiation performance is in good agreement with German fuel performance modeling. • International benchmark exercise expected first particle to fail at ∼13–17% FIMA. • EOL silver release is predicted to be in the percentage range. • EOL cesium and strontium are expected to remain at a low level. - Abstract: Various countries engaged in the development and fabrication of modern HTGR fuel have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under HTGR operating and accident conditions. Verification and validation studies are conducted by code-to-code benchmarking and code-to-experiment comparisons as part of international exercises. The methodology developed in Germany since the 1980s represents valuable and efficient tools to describe fission product release from spherical fuel elements and TRISO fuel performance, respectively, under given conditions. Continued application to new results of irradiation and accident simulation testing demonstrates the appropriateness of the models in terms of a conservative estimation of the source term as part of interactions with HTGR licensing authorities. Within the European irradiation testing program for HTGR fuel and as part of the former EU RAPHAEL project, the HFR-EU1 irradiation experiment explores the potential for high performance of the presently existing German and newly produced Chinese fuel spheres under defined conditions up to high burnups. The fuel irradiation was completed in 2010. Test samples are prepared for further postirradiation examinations (PIE) including heatup simulation testing in the KÜFA-II furnace at the JRC-ITU, Karlsruhe, to be conducted within the on-going ARCHER Project of the European Commission. The paper will describe the application of the German computer models to the HFR-EU1 irradiation test and

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

Science.gov (United States)

Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

1973-01-01

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Power ramp tests of MOX fuel rods. HBWR irradiation with the instrument rig, IFA-591

International Nuclear Information System (INIS)

Ozawa, Takayuki; Abe, Tomoyuki

2006-03-01

Plutonium-uranium mixed oxide (MOX) fuel rods of instrumental rig IFA-591 were ramped in HBWR to study the Advanced Thermal Reactor (ATR) MOX fuel behavior during transient operation and to determine a failure threshold of the MOX fuel rods. Eleven segments were base-irradiated in ATR 'FUGEN' up to 18.4 GWd/t. Zirconium liner claddings were adopted for four segments of them. As the results of non-destructive post irradiation examinations (PIEs) after the base-irradiation and before the ramp tests, no remarkable behavior affecting the integrity of fuel assembly and fuel rod was confirmed. All segments to be used for the ramp tests, which consisted of the multi-step ramp tests and the single-step ramp tests, had instrumentations for in-pile measurements of cladding elongation or plenum pressure, and heated up to the maximum linear power of 58.3-68.4 kW/m without failure. The major results of ramp tests are as follows: There is no difference in PCMI behaviors between two type rods of Zry-2 and Zirconium liner claddings from the in-pile measurements of cladding elongation and plenum pressure. The computations of cladding elongation and inner pressure gave slightly lower elongation and pressure than the in-pile measurements during the ramp-test. However, the cladding relaxation during the power hold was in good agreement, and the fission gas release behavior during cooling down could be evaluated by taking into account the relaxation of contact pressure between pellet and cladding. Although the final power during IFA-591 ramp tests reached the higher linear power than the failure threshold power of UO 2 fuel rods, no indication of fuel failure was observed during the ramp tests. The cladding relaxation due to the creep deformation of the MOX pellets at high temperature could be confirmed at the power steps during the multi-ramp test. The fission gas release due to the emancipation from PCMI stress was observed during the power decreasing. The burn-up dependence could be

Impacto de las unidades a pie de carretera en las interferencias en redes vehiculares

OpenAIRE

Hernández Gañán, Carlos; Reñé Vicente, Sergi; Mata Diaz, Jorge; Alins Delgado, Juan José

2013-01-01

Las redes ad hoc vehiculares (VANETs) han emergido como una tecnolog´ıa clave que posibilita el despliegue diversas aplicaciones. Proporcionar seguridad a los servicios de informaci´on y entretenimiento en estos entornos requiere del uso de unidades a pie de la carretera (RSU) como puertas de acceso a los recursos solicitados. Idealmente, las RSUs deben desplegarse para proporcionar conectividad continua. Si bien esto aumenta la capacidad y cobertura, tambi´en acarrea un ...

Post-irradiation examination of fifteen UO2/PuO2-fuel pins from the experiment DFR-350

International Nuclear Information System (INIS)

Geithoff, D.

1975-06-01

Within the framework of the fuel pin development for a sodium-cooled fast reactor a subassembly containing 77 fuel pins has been irradiated up to 5.65% fima in the Dounreay fast reactor. The pins were prototypes in terms of fuel and cladding material. The fuel consisted of mechanically mixed UO 2 (80%) and PuO 2 (20%) pressed into pellets whereas austenitic steels (W.-No. 1,4961 and 1,4988) were used as cladding material. Furthermore a blanket column of UO 2 pellets and a gas plenum were incorporated in the pin. For irradiation the conditions in a fast breeder were simulated by a linear rod power of 450 W/cm and a maximum cladding temperature of 630 0 C. After the successful completion of the irradiation, the subassembly was dismantled and fifteen pins were selected for a nondestructive and destructive examination. The tests included visual control, measurement of external dimensions, γ-spectroscopy, X-ray radiography, fission gas measurement, ceramography, radiochemical burn-up measurement. The results are presented. The most important results of the examinations seem to be the migration of fission product cesium and the fact that no signs of impending pin failure have been found. Thus the pin specification tested in this experiment is capable of achieving higher burnups under the irradiation conditions described above. (orig./AK) [de

POPULATION ANALYSIS OF THE LOCAL ENDANGERED PŘEŠTICE BLACK-PIED PIG BREED

Directory of Open Access Journals (Sweden)

Emil Krupa

2015-09-01

Full Text Available The pedigree analysis of the local endangered Přeštice Black-Pied pig breed (n=19 289 was performed. Animals born within the period 2012-2014 were assumed as the reference population (n=1 374. The pedigree completeness index reached 100% for four generations back. The 100 % of the genetic pool was explained by 66 ancestors. Although all animals of the reference population were inbred, 57% of them had inbreeding less than five percent. Average inbreeding, co-ancestry coefficient and rate of inbreeding reached 4.93%, 13.48% and 1.29% in reference population, respectively. The effective population size calculated by four different methods varied from 32 to 91 animals in 2014. Average generation interval, average family size for sire and dam parents was 2.5, 17.46 and 6.5 animals, respectively. Total number of founders, effective number of founders, effective number of founders’ genomes and effective number of non-founders genomes reached values 299, 98.05, 21.92 and 28.23 founders, respectively. The average genetic diversity (GD loss was 13.71% in reference population. The GD loss has increased within the last three year period mainly due to the random genetic drift (77.6% and by unequal contribution of founders (22.4%. The Preštice Black-Pied breed is highly endangered with GD loss. Mating of closely related animals has to be prevented in breeding and mating program of this breed.

"A New Kind of Rule": The Subversive Narrator in "Alice's Adventures in Wonderland" and "The Pied Piper of Hamelin."

Science.gov (United States)

Blackburn, William

1986-01-01

Compares "Alice's Adventures in Wonderland" to "The Pied Piper of Hamelin," noting that both: (1) were begun for the amusement of specific children; (2) use a subterranean journey as a device; (3) are critical of social authority; and (4) have problematic endings. (SRT)

Non-destructive test for irradiated fuels using X-ray CT system in hot-laboratory

International Nuclear Information System (INIS)

Kim, Heemoon; Kim, Gil-Soo; Yoo, Boung-Ok; Tahk, Young-Wook; Cho, Moon-Sung; Ahn, Sang-Bok

2015-01-01

To inspect inside of irradiated fuel rod for PIE in hotcell, neutron beam and X-ray have been used. Many hot laboratories in the world have shown the results for NDT by 2-D film data. Currently, computed image processing technology instead of film has been developed and CT was applied to the X-ray and neutron beam system. In this trend, our facility needed to set up X-ray system for irradiated fuel inspection and installed in hotcell with consideration of radiation damage. In this study, X-ray system was tested to be operated with radioactive samples and was performed to inspect fuel rods and observe internal damage and dimensional change. 450kV X-ray CT system was installed in hotcell with modification and tested to check image resolution and radiation damage. The image data were analyzed by 3-D computer software. 8 fuel plates and VHTR rods were inspected and measured internal shape and dimension

AGR-1 Compact 4-1-1 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); van Rooyen, Isabella J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-02-01

Destructive post-irradiation examination was performed on AGR-1 fuel Compact 4-1-1, which was irradiated to a final compact-average burnup of 19.4% FIMA (fissions per initial metal atom) and a time-average, volume-average temperature of 1072°C. The analysis of this compact focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy, electron microscopy, and elemental analysis. Deconsolidation-leach-burn-leach (DLBL) analysis revealed no particles with failed TRISO or failed SiC layers (as indicated by very low uranium inventory in all of the leach solutions). The total fractions of the predicted compact inventories of fission products Ce-144, Cs-134, Cs-137, and Sr-90 that were present in the compact outside of the SiC layers were <2×10^-6, based on DLBL data. The Ag-110m fraction in the compact outside the SiC layers was 3.3×10^-2, indicating appreciable release of silver through the intact coatings and subsequent retention in the OPyC layers or matrix. The Eu-154 fraction was 2.4×10^-4, which is equivalent to the inventory in one average particle, and indicates a small but measurable level of release from the intact coatings. Gamma counting of 61 individual particles indicated no particles with anomalously low fission product retention. The average ratio of measured inventory to calculated inventory was close to a value of 1.0 for several fission product isotopes (Ce-144, Cs-134, and Cs-137), indicating good retention and reasonably good agreement with the predicted inventories. Measured-to-calculated (M/C) activity ratios for fission products Eu-154, Eu-155, Ru-106, Sb

Post-irradiation examination of a failed PHWR fuel bundle of KAPS-2

International Nuclear Information System (INIS)

Mishra, Prerna; Unnikrishnan, K.; Viswanathan, U.K.; Shriwastaw, R.S.; Singh, J.L.; Ouseph, P.M.; Alur, V.D.; Singh, H.N.; Anantharaman, S.; Sah, D.N.

2006-08-01

Detailed post irradiation examination was carried out on a PHWR fuel bundle irradiated at Kakrapar Atomic Power Station unit 2 (KAPS-2). The fuel bundle had failed early in life at a low burnup of 387 MWd/T. Non destructive and destructive examination was carried out to identify the cause of fuel failure. Visual examination and leak testing indicated failure in two fuel pins of the outer ring of the bundle in the form of axial cracks near the end plug location. Ultrasonic testing of the end cap weld indicated presence of lack of fusion type defect in the two fuel pins. No defect was found in other fuel pins of the bundle. Metallographic examination of fuel sections taken from the crack location in the failed fuel pin showed extensive restructuring of fuel. The centre temperature of the fuel had exceeded 1700 degC at this location in the failed fuel pin, whereas fuel centre temperature in the un-failed fuel pin was only about 1300 degC. Severe fuel clad interaction was observed in the failed fuel pin at and near the location of failure but no such interaction was observed in the un-failed fuel pins. Several incipient cracks originating from the inside surface were found in the cladding near failure location in addition to the main through wall crack. The incipient cracks were filled with interaction products and hydride platelets were present at tip of the cracks. It was concluded from the observations that the primary cause of failure was the presence of a part-wall defect in the end cap weld of the fuel pins. These defects opened up during reactor operation leading to steam ingress into the fuel, which caused high fuel centre temperature and severe fuel-cladding interaction resulting in secondary failures. A more stringent inspection and quality control of end plug weld during fabrication using ultrasonic test has been recommended to avoid such failure. (author)

Application of Laser Micro-irradiation for Examination of Single and Double Strand Break Repair in Mammalian Cells.

Science.gov (United States)

Holton, Nathaniel W; Andrews, Joel F; Gassman, Natalie R

2017-09-05

Highly coordinated DNA repair pathways exist to detect, excise and replace damaged DNA bases, and coordinate repair of DNA strand breaks. While molecular biology techniques have clarified structure, enzymatic functions, and kinetics of repair proteins, there is still a need to understand how repair is coordinated within the nucleus. Laser micro-irradiation offers a powerful tool for inducing DNA damage and monitoring the recruitment of repair proteins. Induction of DNA damage by laser micro-irradiation can occur with a range of wavelengths, and users can reliably induce single strand breaks, base lesions and double strand breaks with a range of doses. Here, laser micro-irradiation is used to examine repair of single and double strand breaks induced by two common confocal laser wavelengths, 355 nm and 405 nm. Further, proper characterization of the applied laser dose for inducing specific damage mixtures is described, so users can reproducibly perform laser micro-irradiation data acquisition and analysis.

Prevalencia y grado de pie plano según estado nutricional en niños escolares de las instituciones educativas Santa Isabel de Hungría y Santa Teresita de Jesús. Casma. 2010

OpenAIRE

Sánchez Reyna, Luis Fernando

2016-01-01

El objetivo del estudio fue establecer la prevalencia y grado de pie plano en relación al estado nutricional en niños escolares matriculados en las Instituciones Educativas Santa Isabel de Hungría y Santa Teresita de Jesús de la ciudad de Casma. Este estudio cuantitativo, descriptivo y transversal evaluó a 321 niños escolares durante el año 2010. Se clasificó en niños eutróficos, con sobrepeso y obesidad según las tablas del IMC para la edad de la OMS, y en niños con pie normal o pie plano de...

Rapportering om going concern i den nye revisionspåtegning – særlige forhold for PIE-virksomheder

DEFF Research Database (Denmark)

Seehausen, Jesper

2017-01-01

Artiklen omhandler særlige forhold for PIE-virksomheder, dvs. virksomheder af interesse for offentligheden, vedrørende going concern. Artikel skal læses i sammenhæng med min generelle artikel vedr. rapportering om going concern i den nye revisionspåtegning, idet denne artikel også er relevant...

Post irradiation examination of type 316 stainless steels for in-pile Oarai water loop No.2 (OWL-2)

International Nuclear Information System (INIS)

Shibata, Akira; Kimura, Tadashi; Nagata, Hiroshi; Aoyama, Masashi; Kanno, Masaru; Ohmi, Masao

2010-11-01

The Oarai water loop No.2 (OWL-2) was installed in JMTR in 1972 for the purpose of irradiation experiments of fuel element and component material for light water reactors. Type 316 stainless steels (SSs) were used for tube material of OWL-2 in the reactor. But data of mechanical properties of highly irradiated Type 316 SSs has been insufficient since OWL-2 was installed. Therefore surveillance tests of type 316 SSs which were irradiated up to 3.4x10 25 n/m 2 in fast neutron fluence (>1 MeV) were performed. Meanwhile type 316 stainless steel (SS) is widely used in JMTR such as other irradiation apparatus and irradiation capsule, and additional data of type 316 SSs irradiated higher is required. Therefore post irradiation examinations of surveillance specimens made of type 316 SSs which were irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence were performed and reported in this paper. In this result of surveillance tests of type 316 SSs irradiated up to 1.0x10 26 n/m 2 , tensile strength increase with increase of Neutron fluence and total elongation decreased with increase of Neutron fluence compared to unirradiated specimens and specimens irradiated up to 3.4x10 25 n/m 2 . This tendency has good agreement with results of 10 24 - 10 25 n/m 2 in fast neutron fluence. More than 37% in total elongation was confirmed in all test conditions. It was confirmed that type 316 SS irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence has enough ductility as structure material. (author)

Irradiation behaviour of a tritium breeding material, γ-LiAlO 2- results of two in-pile experiments: ALICE I and ALICE II

Science.gov (United States)

Botter, F.; Rasneur, B.; Roth, E.

1988-11-01

γ-LiAlO 2 has been studied at CEA as potential breeder material for fusion reactors within the scope of the EEC fusion technology program. Radiation damage was investigated by irradiating unclad aluminate samples in the core of the OSIRIS reactor at Saclay. As part of the international breeder material comparison program named BEATRIX, US samples were irradiated along with those prepared in Saclay; samples of natural 6Li content and 96% enriched ones were irradiated. Shapes were chosen to enable postirradiation examinations (PIE), and microstructures were optimized for tritium release. The ALICE 1 experiment was carried out during 25.7 full power days (FPD), ALICE II lasted 36.3 FPD. Temperatures ranged from 400 to 600°C in the first, from 750 to 850°C in the second ALICE irradiation (sample core temperatures). In both cases the maximum flux on the samples was 2.1 × 10 18n m -2 s -1 fast, and 0.7 × 10 18n m -2 s -2 thermal Power dissipated was up to 100 W/cm 3, higher than the average in most reactor blanket designs by a factor 3 to 10, thus enabling the highest burn-ups to correspond to more than two years of possible operation in a full-scale reactor. In the lower temperature range of irradiation no significant damage was observed. In the higher one shrinkage due to sintering was induced. Whatever the microstructure, the flux and temperature, all samples (but one) not exceeding 5 mm diameter and length were mechanically intact. Above those dimensions cracking, which can be assigned to excessive thermal stress, could be observed. Given anticipated operating conditions of blankets being designed, the behaviour of γ-LiAlO 2 under irradiation is that of a very promising material.

Biennial activity report of the Division for PIE and NDT Development [for years] 1989-1990

International Nuclear Information System (INIS)

Babu Rao, C.; Shyamsunder, M.T.; Murugan, S.; Ramabathiran, A.; Rajagopalan, C.; Bhattacharya, D.K.; Kumar, P.V.; Kalyanasundaram, P.; Kasiviswanathan, K.V.

1992-01-01

This is the first biennial report of the Division for Post-Irradiation Examination and Non-Destructive Testing (DPEND) of the Indira Gandhi Centre for Atomic Research, Kalpakkam. It covers the research and development (R and D) activities of the Division during 1989 and 1990. The R and D activities of the Division are both multidisciplinary and interdisciplinary. The principal areas of R and D of DPEND are post-irradiation examination of fuel and structural materials, irradiation experiments, non-destructive evaluation for material characterisation, quality assurance and inservice inspection, failure investigations, remote technology etc. These R and D activities are reported in the form of extended summaries arranged under the headings: research and development in non-destructive evaluation, material characterisation, instrumentation, development of techniques, software development, mechanical equipment and systems (development and commissioning), inspections, and failure analysis. Some of the highlights of these activities are : (1)setting up of a radio metallurgy laboratory with hot cells with recirculating nitrogen gas environment to facilitate post-irradiation examination of the carbide fuel of the Fast Breeder Test Reactor. (FBTR), (2) development of the remote milling and drilling machine and development of pressurised capsules for irradiation creep measurement experiments in FBTR, (b) fabrication of non-destructive testing (NDT) reference standards with controlled defects for use in Nuclear Fuel Complex, Hyderabad. There are a number of appendices listing publications, conference paper, invited talks, internal reports etc. by the scientists of the Division and also awards/prizes won by the scientists. A staff-chart of the Division is also given. (M.G.B.)

Postirradiation examination of JOYO MK-II control rod (CRM601). Irradiation performance of shroud type absorber pin

International Nuclear Information System (INIS)

Tanaka, Kosuke; Kikuchi, Shin; Katsuyama, Kozo; Nagamine, Tsuyoshi; Mitsugi, Takeshi; Uto, Manabu; Tatebe, Kazuaki; Onose, Shoji; Maruyama, Tadashi

1998-10-01

This paper describes the results of postirradiation examination and analysis by CORAL code for irradiation performance of CRM601 control rod, which was the 6th reloaded control rod with shroud type absorber pins for use in JOYO MK-II core. The detailed visual examination indicated that there was no cladding breach in absorber pins. However, sodium ingress from the vent tube was observed in four absorber pins among seven pins. While a remarkable oval deformation occurred in cladding tube of helium bonded absorber pins, a little or no diametral change was observed in the absorber pins in which sodium ingress took place. From metallurgical observations and the analysis by CORAL code, it was estimated that the shroud tube installed in helium bonded absorber pins were irradiated at 720degC, and those in sodium bonded absorber pins were irradiated at 420degC. It was confirmed that diametral change of cladding depended on the initial gap between shroud and cladding tube. The results of present investigation indicate that it is desirable to use the materials with low thermal expansion coefficient for shroud tubes, and that sodium bonded absorber pins were advantageous for obtaining long life control rods. (author)

Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

Energy Technology Data Exchange (ETDEWEB)

Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D. [Chalk River Labs., Ontario (Canada)

1997-08-01

Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

TEM examination of the effect of post-irradiation annealing on 7.7 dpa AISI 304 stainless steel

International Nuclear Information System (INIS)

Karlsen, W.; Ivanchenko, M.; Pakarinen, J.; Karlsen, T.

2015-01-01

Stainless steels exposed to neutron irradiation during service in light water reactors (LWR) can become susceptible to intergranular cracking, referred to as irradiation assisted stress corrosion cracking (IASCC). Analytical transmission electron microscopy (ATEM) was used to examine the effect of post-irradiation annealing (PIA) on radiation-induced segregation (RIS) at the grain boundaries of 7.7 dpa AISI 304 stainless steel. The grain boundary profiles and the irradiation damage were analysed in the as-irradiated state and after PIA of 6 hours at 500 C. degrees and after 25 hours at 500 C. degrees and 550 C. degrees by using transmission electron microscopy (TEM). As a main conclusion from the TEM examinations, the effects of PIA were found to be relatively small after only 6 hours, while after 25 hours of PIA at both 500 and 550 C. degrees, RIS was almost recovered and only marginal deviation in chemical composition could be found near the GB. The as-irradiated state showed extreme RIS values of Si 4.9 wt%, Cr 14.7 wt%, Ni 23.4 wt%, and P 1.4 wt%., while upon PIA for 6 hours the extreme values for RIS were Si 3.9 wt%, Cr 16.0 wt%, Ni 21 wt%, and P 0.9 wt%. After 6 hours annealing at 500 C. dislocation loops start to grow, while dislocation density remains of the same order of magnitude. After annealing for 25 hours at 500 C. degrees the average size of dislocation loops remains nearly the same, while dislocation density was reduced almost by one fold. In the areas where dislocation density was found to be the lowest some features, which can most likely be attributed to stacking fault tetrahedral (SFT) were found. Annealing at even higher temperature (550 C.) affected the average size of the dislocation loops, making them almost twice as large as well as resulting in a very broad distribution of dislocation sizes. Density of dislocations is also reduced by one fold in comparison to the as irradiated condition and leads to formation of SFTs, which could be

Estratificación de riesgo en pie diabético Risk stratification in diabetic foot

Directory of Open Access Journals (Sweden)

Héctor González de la Torre

2010-12-01

Full Text Available El pie diabético es una complicación crónica de la diabetes mellitus y constituye una problemática sanitaria al conllevar un enorme gasto económico a los servicios sanitarios y provocar a las personas que lo padecen y a su entorno familiar enormes repercusiones emocionales. La educación sanitaria en el cuidado de los pies de los diabéticos, así como la cuantificación del riesgo son aspectos que el enfermero debe incorporar en los planes de cuidados de los pacientes diabéticos. La detección de los posibles factores de riesgo es un paso previo imprescindible para estos cometidos. Este estudio tiene como finalidad analizar la presencia de determinados factores de riesgo en pie diabético en la población diabética del Centro de Salud de Triana, con el objetivo de poder realizar, así, una estratificación del riesgo. Para ello, se optó por la realización de un estudio observacional descriptivo. La muestra analizada estaba constituida por 96 sujetos diabéticos pertenecientes a dicho centro de salud. Para la recogida de datos se utilizó un sistema de tres niveles consistente en entrevista, exploración física y consulta de registros disponibles (a través de las historias clínicas de los sujetos. Los datos se registraron en un formulario de recogida de datos para pie diabético. Se determinó la existencia de factores de riesgo, analizando las relaciones entre éstos y permitiendo la estratificación del riesgo en la población estudiada.The diabetic foot is a chronic complication of diabetes mellitus and constitutes a sanitary problem which causes not only enormous expenses for the sanitary system but also provokes emotional repercussions for the person who suffers the illness and for the family. The sanitary education in taking care of the feet of diabetics and the identification of the health risk are both aspects that the nurse must include in the healthcare planning of diabetic patients. The detection of the possible risk

The YouTube Project: “The PewDiePie versus The Wall Street Journal Scandal – an in-depth analysis of PewDiePie’s online audience”

OpenAIRE

Jensen, Natalie Madeleine; Rasmussen, Anna Melanie Bæckmann; Cillo, Dominic

2017-01-01

In this dissertation, the aim is to investigate the multi-faceted events and belonging themes within The PewDiePie versus The Wall Street Journal Scandal – henceforth described as The PewDiePie Scandal. Here, the main goal within this dissertation is to answer questions in regards to PewDiePie’s online audience, how they perceive him and the relationship between the YouTuber and the viewership. Furthermore, the investigation will delve into matters of the major support provided by his audienc...

MICRO/NANO-STRUCTURAL EXAMINATION AND FISSION PRODUCT IDENTIFICATION IN NEUTRON IRRADIATED AGR-1 TRISO FUEL

Energy Technology Data Exchange (ETDEWEB)

van Rooyen, I. J.; Lillo, T. M.; Wen, H. M.; Hill, C. M.; Holesinger, T. G.; Wu, Y. Q.; Aguiara, J. A.

2016-11-01

Advanced microscopic and microanalysis techniques were developed and applied to study irradiation effects and fission product behavior in selected low-enriched uranium oxide/uranium carbide TRISO-coated particles from fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA. Although no TRISO coating failures were detected during the irradiation, the fraction of Ag-110m retained in individual particles often varied considerably within a single compact and at the capsule level. At the capsule level Ag-110m release fractions ranged from 1.2 to 38% and within a single compact, silver release from individual particles often spanned a range that extended from 100% retention to nearly 100% release. In this paper, selected irradiated particles from Baseline, Variant 1 and Variant 3 type fueled TRISO coated particles were examined using Scanning Electron Microscopy, Atom Probe Tomography; Electron Energy Loss Spectroscopy; Precession Electron Diffraction, Transmission Electron Microscopy, Scanning Transmission Electron Microscopy (STEM), High Resolution Electron Microscopy (HRTEM) examinations and Electron Probe Micro-Analyzer. Particle selection in this study allowed for comparison of the fission product distribution with Ag retention, fuel type and irradiation level. Nano sized Ag-containing features were predominantly identified in SiC grain boundaries and/or triple points in contrast with only two sitings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). STEM and HRTEM analysis showed evidence of Ag and Pd co-existence in some cases and it was found that fission product precipitates can consist of multiple or single phases. STEM analysis also showed differences in precipitate compositions between Baseline and Variant 3 fuels. A higher density of fission product precipitate clusters were identified in the SiC layer in particles from the Variant 3 compact compared with the Variant 1 compact. Trend analysis shows

Song degradation in the hole-nesting pied flycatcher Ficedula hypoleuca: Implications for polyterritorial behaviour in contrasting habitat-types

DEFF Research Database (Denmark)

Lampe, Helene M.; Larsen, Ole Næsbye; Pedersen, Simon Boel

2007-01-01

In the hole-nesting pied flycatcher, Ficedula hypoleuca, a male may become polyterritorial after attracting a primary female. However, the female may recognize her mate's song and attack other females that associate with him. Differences in sound degradation amongst different habitats and within ...

The treatment of active waste from a PIE facility

International Nuclear Information System (INIS)

Turier, C.A.; Kerswell, A.G.

1978-09-01

The types of radioactive waste produced in the post irradiation examination of nuclear fuel elements from several classes of reactor are described. Other radioactive wastes may be produced in cave facilities as a result of contamination of the equipment. The methods of disposal of all types of waste are considered, together with methods to improve the operation of the caves. The training of cave operators, and the use of method study to collect information in cave operations are considered also. (U.K.)

Qualification of the on-line power determination of fuel elements in irradiation devices in the BR2 reactor

International Nuclear Information System (INIS)

Vermeeren, L.; Dekeyser, J.; Gouat, P.; Kalcheva, S.; Koonen, E.; Kuzminov, V.; Verwimp, A.; Weber, M.

2005-01-01

Fuel irradiation tests require an on-line monitoring of the fuel power. In the BR2 reactor, this is performed by continuously measuring the enthalpy change in the coolant of the irradiation device and complementing this information with data on power losses, heating of structure parts and spatial power profiles from mock-up test experiments and from calculations. Since a few years Monte Carlo codes (MCNP) are used, describing the BR2 core in great detail for every reactor cycle with its specific core load, yielding not only reliable relative values, but also calculated absolute local power values in agreement with data from PIE analyses. Several methods were conceived to combine the experimental and calculated data for the on-line calculation of the local linear power in the fuel elements; their internal consistency and the consistency with gamma spectroscopy data and data from radiochemical fission product analysis was checked. The data show that fuel irradiations in BR2 can be performed in a well-controlled way, with an accurate and reliable on-line follow-up of the fuel power. (author)

The PIES2012 Code for Calculating 3D Equilibria with Islands and Stochastic Regions

Science.gov (United States)

Monticello, Donald; Reiman, Allan; Raburn, Daniel

2013-10-01

We have made major modifications to the PIES 3D equilibrium code to produce a new version, PIES2012. The new version uses an adaptive radial grid for calculating equilibrium currents. A subset of the flux surfaces conform closely to island separatrices, providing an accurate treatment of the effects driving the neoclassical tearing mode. There is now a set of grid surfaces that conform to the flux surfaces in the interiors of the islands, allowing the proper treatment of the current profiles in the islands, which play an important role in tearing phenomena. We have verified that we can introduce appropriate current profiles in the islands to suppress their growth, allowing us to simulate situations where islands are allowed to grow at some rational surfaces but not others. Placement of grid surfaces between islands is guided by the locations of high order fixed points, allowing us to avoid spectral polution and providing a more robust, and smoother convergence of the code. The code now has an option for turning on a vertical magnetic field to fix the position of the magnetic axis, which models the horizontal feedback positioning of a tokamak plasma. The code has a new option for using a Jacobian-Free Newton Krylov scheme for convergence. The code now also contains a model that properly handles stochastic regions with nonzero pressure gradients. Work supported by DOE contract DE-AC02-09CH11466.

Post Irradiation Examination Results of the NT-02 Graphite Fins NUMI Target

Energy Technology Data Exchange (ETDEWEB)

Ammigan, K. [Fermilab; Hurh, P. [Fermilab; Sidorov, V. [Fermilab; Zwaska, R. [Fermilab; Asner, D. M. [PNL, Richland; Casella, Casella,A.M [PNL, Richland; Edwards, D. J. [PNL, Richland; Schemer-Kohrn, A. L. [PNL, Richland; Senor, D. J. [PNL, Richland

2017-02-10

The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW, and saw an integrated total proton on target of 6.1 1020. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-termstructural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

Phosphorus sources (P plus filter pie with or without azotofos on the available P in the soil

Directory of Open Access Journals (Sweden)

Maikel Abreu Jiménez

2014-04-01

Full Text Available The objective of the investigation was to evaluate the effect of four phosphorus sources plus filter pie with or without the biofertilizer Azotofos on the available phosphorus in the soil at different moments after the treatment. An experiment in factorial design 4(2+1 was established, being the four phosphorus sources: rock phosphate, natural phosphate, triple phosphate and Cuban phosphoric rock; two sources of the organic compound to base filter cake enriched with Azotofos microorganisms, only filter cake (without enrichment and a control treatment (without filter pie, neither Azotofos, with three repetitions. The evaluations of the tenor of available P (Bray-2 were carried out at the 30, 60, 90, 120 and 150 days after the installation of the experiment. The tenor of P (Bray-2 was influenced by the sources of P and the enrichment with biofertilizantes (factorial increasing the tenor of available P in front of the control. The triple superphosphate promoted the higher tenors in P in the soil at 60 and 90 days after its application, independently of the presence or not of the organic compound enriched with P solubilizing microorganisms, although this effect didn’t stay stable at the time.

Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W

International Nuclear Information System (INIS)

Rao, Y.F.; Manzer, A.M.

2005-01-01

This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)

Preposition stranding versus pied-piping: Negative Shift of prepositional complements in dialects of Faroese

Directory of Open Access Journals (Sweden)

Eva Engels

2009-01-01

         These asymmetries will be accounted for within Fox and Pesetsky's (2003, 2005 cyclic linearization model, which requires non-string-vacuous movement to proceed through the left edge of Spell-out domains, deriving cross-linguistic variation as to Negative Shift from differences in the availability of these left-edge positions. Thereby, pied-piping is considered a last resort strategy, possible only if the prepositional complement cannot undergo Negative Shift on its own due to the unavailability of the relevant left-edge position.

Preliminary examination of induced radioactivity in pepper by 10 MeV electron irradiation

International Nuclear Information System (INIS)

Katayama, Tadashi; Furuta, Masakazu; Sibata, Setsuko; Ito, Norio; Mizohata, Akira; Matsunami, Tadao; Toratani, Hirokazu; Takeda, Atsuhiko.

1991-01-01

β-ray measurement was performed on 10 MeV electron-irradiated black pepper and white pepper with liquid scintillation counter in order to reconfirm the wholesomeness of irradiated foods and present unambiguous data to general consumers concerning about the induced radioactivity in the irradiated foods. In irradiated black pepper no radioactivity other than from natural source, un-irradiated one, was detected. But in irradiated white pepper, it was suggested that induced radioactivity might be detected if the detection method was more improved. (author)

Advanced Test Reactor National Scientific User Facility Partnerships

International Nuclear Information System (INIS)

Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

2012-01-01

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

Power-to-melt evaluation of fresh mixed-oxide fast reactor fuel. Technical improvements of the post-irradiation-experiment and the evaluation of the results for the power-to-melt test PTM-2 in 'JOYO'

International Nuclear Information System (INIS)

Yamamoto, Kazuya; Kushida, Naoya; Koizumi, Atsuhiro

1999-11-01

The second Power-To-Melt (PTM) test, PTM-2, was performed in the experimental fast reactor 'JOYO'. All of the twenty-four fuel pins of the irradiation vehicle, B5D-2, for the PTM-2 test, were provided for post-irradiation-experiment (PIE) to evaluate the PTM values. In this study, the PIE technique for PTM test was established and the PTM results were evaluated. The findings are as follows: The maximum fuel-melting ratio on the transverse section was 10.7%, and was within the limit of fuel-melting in this PTM test enough. Unexpected fuel-melting amount to a ratio of 11.8% was found at ∼24 mm below the peak power elevation in a test fuel pin. It is possible that this arose from secondary fuel-melting. Combination of metallographical observation with X-ray microanalysis of plutonium distribution was very effective for the identification of once-molten fuel zone. The PTM evaluation suggested that dependence of the PTM on the fuel pellet density was stronger than that of previous foreign PTM tests, while the dependence on the pellet-cladding gap and the oxygen-to-metal ratio was indistinctly. The dependence on the cladding temperature and the fill gas composition was not shown as well. (author)

Status and Update of the RaDIATE Collaboration R&D Program

Energy Technology Data Exchange (ETDEWEB)

Ammigan, K. [Fermilab; Hurh, P. [Fermilab

2017-07-01

The Radiation Damage In Accelerator Target Environments (RaDIATE) collaboration was founded in 2012 and currently consists of over 50 participants and 11 institutions globally. Due to the increasing power of future proton accelerator sources in target facilities, there is a critical need to further understand the physical and thermo-mechanical radiation response of target facility materials. Thus, the primary objective of the RaDIATE collaboration is to draw on existing expertise in the nuclear materials and accelerator targets fields to generate new and useful materials data for application within the accelerator and fission/fusion communities. Current research activities of the collaboration include post irradiation examination (PIE) of decommissioned components from existing beamlines such as the NuMI beryllium beam window and graphite NT-02 target material. PIE of these components includes advanced microstructural analyses (SEM/TEM, EBSD, EDS) and micro-mechanics technique such as nano-indentation, to help characterize any microstructural radiation damage incurred during operation. New irradiation campaigns of various candidate materials at both low and high energy beam facilities are also being pursued. Beryllium helium implantation studies at the University of Surrey as well as high energy proton irradiation of various materials at Brookhaven National Laboratory’s BLIP facility have been initiated. The program also extends to beam-induced thermal shock experiments using high intensity beam pulses at CERN’s HiRadMat facility, followed by advanced PIE activities to evaluate thermal shock resistance of the materials. Preliminary results from ongoing research activities, as well as the future plans of the RaDIATE collaboration R&D program will be discussed.

Post-Irradiation Examination of Fuel Pin R54-F20A, Irradiated in a NaK Environment. RCN Report

International Nuclear Information System (INIS)

Kwast, H.

1972-12-01

Fuel pin R54-F20A has been irradiated in a NaK-environment. Temperature measurements in the NaK were carried out at average linear fission powers of 552 and 825 W/cm respectively. A maximum average canning temperature of 920°C was reached. The fuel pin was irradiated for about 50 minutes at the maximum irradiation conditions, while the total irradiation time was two hours. The irradiation had to be broken off before the end condition was reached because of malfunctioning of the fuelfailure detection system. No power peaking did occur at the upper and lower interfaces between the 50%-enriched UO 2 - and the natural UO 2 + 8 w/o UB 4 pellet. About 35% of the fuel has molten, but the fuel pin did not fail. The irradiation has been carried out in the Poolside Facility (PSF) of the High Flux Reactor (HFR) at Petten. (author)

Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

Energy Technology Data Exchange (ETDEWEB)

Van Rooyen, Isabella Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riesterer, Jessica Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, Brandon Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott Arden [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2012-12-01

The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compact 6-3-2 has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm

Consumer opinions in Argentina on food irradiation: irradiated onions

International Nuclear Information System (INIS)

Curzio, O.A.; Croci, C.A.

1998-01-01

Two surveys were carried out in Buenos Aires of consumer attitudes towards irradiated onions [no data given]. The first investigated the general level of consumer knowledge concerning food irradiation, whilst the second (which covered consumers who had actually bought irradiated onions) examined reasons for purchase and consumer satisfaction. Results reveal that more than 90% of consumers surveyed had a very limited knowledge of food irradiation

Safer prepared meals for immunocompromised patients and the general consumer by gamma irradiation

International Nuclear Information System (INIS)

Narvaiz, P.; Horak, C.; Campos, M.; Veronesi, P.; Cossani, E.; Lound, L.; Gasparovich, A.; Liendo, G.; Hovsepian, J.; Mengoni, G.

2009-01-01

Ready-to-eat meals are commonplace nowadays in urban life. Many of them, minimally processed, could convey foodborne pathogens likely to cause diseases in the consumer, which is of concern to the normal population and even more so to immunocompromised patients. The feasibility of attaining microbiological decontamination at pasteurization levels of such foods by gamma irradiation was studied. Typical Argentine dishes were chosen after market surveys: cannelloni in tomato sauce, tomato and carrot salad with boiled egg, empanada, fruit salad in gelatin jelly with white cheese, ham and cheese sandwich, chicken and vegetable pie, custard, and bread pudding, in different packaging. Microbiological profiles of the meals were obtained and challenge tests with Listeria innocua or Salmonella enteritidis were performed to determine the minimum radiation dose to be applied in each food so as to attain a 6 log cycle reduction of pathogen counts. Preliminary sensory evaluations, out of panel, were carried out to determine possible evident sensory alterations due to the irradiation treatment. Then, a greater number of samples were irradiated at the minimum radiation dose and at a maximum dose equal to or less than twice the minimum, at the 60 Co semi-industrial facility of the Ezeiza Atomic Center. Microbiological and sensory analyses by a consumer panel were performed on control and irradiated samples throughout storage life at refrigeration temperatures. A whole irradiated lunch, composed of salad, empanadas and fruit salad, was sampled by 44 immunocompromised patients at a hospital. The composition and adequacy of this lunch was designed by nutritionists. Results showed that it was feasible to attain the proposed decontamination goal without significantly impairing the sensory quality. Shelf life was almost tripled in irradiated samples. Immunocompromised patients enjoyed the irradiated lunch and requested it and other dishes to be made commercially available. (author)

Evaluation of the characteristics of uranium and plutonium Mixed Oxide (MOX) fuel

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

MOX fuel irradiation test up to high burnup has been performed for five years. Irradiation test of MOX fuel having high plutonium content has also been performed from JFY 2007 and it still continues. A lot of irradiation data have been obtained through these tests. The activities done in JFY 2012 are mainly focused on Post Irradiation Examination (PIE) data analysis concerning thermal property change and fission gas release. In the former work thermal conductivity degradation due to burnup is examined and in the latter work the dependence of fission gas release mechanism on fuel pellet microstructure is examined. This report mainly covers the result of analysis. It is found that thermal conductivity degradation of MOX fuel due to burnup is less than that of UO{sub 2} fuel and that fission gas release mechanism of high enriched fissile zone (so called Pu spot) is much different from that of low enriched fissile zone (so called Matrix). (author)

Analysis of recent post irradiation tests by Japanese and French burnup analysis code systems

International Nuclear Information System (INIS)

Iwasaki, Tomohiko; Hiraizumi, Hiroaki; Youinou, Gilles

2002-01-01

Benchmark problem based on Japanese Post Irradiation Experiment (PIE) data was analyzed by Japanese burnup analysis code and French one under the cooperative research program between the Japanese University Association (JUA) in Japan and Commissariat a l'Enegie Atomique (CEA) in France. Significant discrepancies over 10% were found between the Japanese and French results for 238 Pu, 243 Am, 244 Cm, 125 Sb, 154 Eu, 134 Cs and 144 Ce. It is supposed that the difference of C/E for 243 Am and 244 Cm between Japanese results and French ones is due to the (n,gamma) reaction of 242m Am. For 125 Sb and 154 Eu, the C/E values are improved by using new cross section and fission yield libraries. (author)

Degradation of song in a species using nesting holes: the Pied Flycatcher Ficedula hypoleuca

Directory of Open Access Journals (Sweden)

Helene M. Lampe

2004-06-01

Full Text Available The habitat, but also the nest hole of a hole-nesting species, will degrade the song during transmission. We investigated how the sounds degrade in a sound transmission experiment with the song of the Pied Flycatcher Ficedula hypoleuca (Muscicapidae. Ten different song elements were transmitted to microphones placed inside and outside a nest box. On average, song degradation was much greater inside than outside the nest boxes, especially with respect to excess attenuation and blurring of the song elements. Being inside a nest box therefore strongly reduces a Pied Flycatcher's possibility of detecting and recognizing songs or eavesdropping on singing interactions.O habitat, mas também a cavidade do ninho de uma espécie nidificando em ocos, provoca uma degradação do som durante sua transmissão. Pesquisamos como os sons são degradados num experimento de transmissão sonora usando o canto do pássaro Ficedula hypoleuca (Muscicapidae. Dez elementos diferentes do canto foram enviados para microfones colocados dentro e fora de uma caixa de nidificação. Em média, a degradação do canto foi muito maior dentro do que fora do ninho, particularmente no que diz respeito ao excesso de atenuação e à borrada dos elementos do canto. Estar dentro de uma cavidade de nidificação, portanto, reduz fortemente a capacidade da espécie de detectar e reconhecer cantos ou de interceptar interações vocais.

Annual report of the JMTR 1992

International Nuclear Information System (INIS)

1993-12-01

The Japan Materials Testing Reactor (JMTR) is capable of a wide variety of irradiations under controlled environmental conditions. The JMTR has been used to perform engineering tests on nuclear fuels and components, for general materials testing, and for radioisotope production since 1970. The JMTR was operated 5 cycles in fiscal 1992, and extensive irradiation tests including power ramping tests were performed. The development of the latest irradiation and post irradiation examination (PIE) technology, such as the thermocouple instrumentation technique to an irradiated fuel and the measurement technique of irradiation embrittlement with non-destructive inspection, is being carried out in the JMTR. The investigation of irradiation behaviors for the blanket materials in the fusion reactor is also being carried out. The core conversion to low-enrichment uranium (LEU) fuel in the JMTR will be fully completed in January 1994. (author)

Quality of pies obtained of the extraction of sunflower and crambe oil for biodiesel under different temperatures; Qualidade das tortas obtidas na extracao do oleo de girassol e crambe para biodiesel sob diferentes temperaturas

Energy Technology Data Exchange (ETDEWEB)

Bischoff, Tabata Zingano; Prado, Naimara Vieira do; Coelho, Silvia Renata Machado; Bischoff, Teodato Zingano [Universidade Estadual do Oeste do Parana (UNIOESTE), PR (Brazil)], emails: tabatazbi@yahoo.com.br, nai_mara@hotmail.com, srmcoelho@unioeste.br, teodato_zb@hotmail.com; Figueiredo, Paulo Roberto Abreu de [Instituto Agronomico do Parana (IAPAR), Londrina, PR (Brazil)

2011-07-01

Sunflower and crambe are viable alternatives to ethanol. Used in food production, the sunflower crop has economic, rustic and does not require specialized machinery. The crambe oil is concentrated enough, it is feasible to farm and is not edible. The extraction efficiencies of these oils gives to the heating and moisture content of grain. The aim of this study was to evaluate temperatures of oil extraction in sunflower and crambe as pies obtained. For oil extraction and acquisition of pies, we used a high pressure extruder press. The sunflower seed and crambe were extruded without heating and at temperatures of 50 and 70 deg C with three replicates for each treatment, estimated to be the pie, the amount of fat, protein and moisture. Increasing the temperature, we observed a reduction of moisture in sunflower cake, but this did not occur in seeds of crambe. For sunflower, the higher levels of lipids occurred for pie without heating, and for crambe, the higher levels of lipids occurred at 50 deg C. As for the protein, was found in both cultures, the highest level at 70 deg C. It is concluded that the extraction temperature alters the composition of oilseed cakes. (author)

Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

Energy Technology Data Exchange (ETDEWEB)

Van Rooyen, Isabella Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riesterer, Jessica Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miller, Brandon Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott Arden [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2012-12-01

The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO₂ kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compact 6-3-2 has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm

Thermophysical instruments for non-destructive examination of tightness and internal gas pressure or irradiated power reactor fuel rods

International Nuclear Information System (INIS)

Pastoushin, V.V.; Novikov, A.Yu.; Bibilashvili, Yu.K.

1998-01-01

The developed thermophysical method and technical instruments for non-destructive leak-tightness and gas pressure inspection inside irradiated power reactor fuel rods and FAs under poolside and hot cell conditions are described. The method of gas pressure measuring based on the examination of parameters of thermal convection that aroused in gas volume of rod plenum by special technical instruments. The developed method and technique allows accurate value determination of not only one of the main critical rod parameters, namely total internal gas pressure, that forms rod mean life in the reactor core, but also the partial pressure of every main constituent of gaseous mixture inside irradiated fuel rod, that provides the feasibility of authentic and reliable leak-tightness detection. The described techniques were experimentally checked during the examination of all types power reactor fuel rods existing in Russia (WWER, BN, RBMK) and could form the basis for new technique development for non-destructive examination of PWR (and other) type rods and FAs having gas plenum filled with spring or another elements of design. (author)

Facilities for post-irradiation examination of experimental fuel elements at Chalk River Nuclear Laboratories

International Nuclear Information System (INIS)

Mizzan, E.; Chenier, R.J.

1979-10-01

Expansion of post-irradiation facilities at the Chalk River Nuclear Laboratories and steady improvement in hot-cell techniques and equipment are providing more support to Canada's reactor fuel development program. The hot-cell facility primarily used for examination of experimental fuels averages a quarterly throughput of 40 elements and 110 metallographic specimens. New developments in ultrasonic testing, metallographic sample preparation, active storage, active waste filtration, and fissile accountability are coming into use to increase the efficiency and safety of hot-cell operations. (author)

Fernando González: caminando por su novela Viaje a pie

Directory of Open Access Journals (Sweden)

FRANCISCO LUIS GIRALDO GUTIÉRREZ

2017-01-01

Full Text Available Fernando González, ícono de la tradición literaria antioqueña, ha hecho que hoy en día las profundas y polémicas ideas sobre la realidad de su época lo presenten como un escritor con alta sensibilidad social, política y un constante renegar religioso. “Caminar” por la obra Viaje a pie , es precisamente eso: recorrer los senderos que él transitó en un tiempo. En el presente artículo se busca conocer la concepción que el autor tiene sobre la geografía, el amor, el desamor, la familia, la política y la religión, procurando así una caracterización del pensador envigadeño, la contextualización de los hechos y escenarios presentes en la obra referenciada.

Blind prediction exercise on modeling of PHWR fuel at extended burnup

International Nuclear Information System (INIS)

Sah, D.N.; Viswanathan, U.K.; Viswanadham, C.S.; Unnikrishnan, K.; Rath, B.N.

2008-01-01

A blind prediction exercise was organised on Indian Pressurised Heavy Water Reactor (PHWR) fuel to investigate the predictive capability of existing codes for their application at extended burnup and to identify areas of improvement. The blind problem for this exercise was based on a PHWR fuel bundle irradiated in Kakrapar Atomic Power Station-I (KAPS-I) up to about 15 000 MWd/tU and subjected to detailed post-irradiation examination (PIE) in the hot cells facility at BARC. Eleven computer codes from seven countries participated in this exercise. The participants provided blind predictions of fuel temperature, fission gas release, internal gas pressure and other performance parameters for the fuel pins. The predictions were compared with the experimental PIE data which included fuel temperature derived from fuel restructuring, fission gas release measured by fuel pin puncturing, internal gas pressure in pin, cladding oxidation and fuel microstructural data. The details of the blind problem and an analysis of the results of blind predictions by the codes vis-a-vis measured data are provided in this paper

Irradiation and pregnancy

Energy Technology Data Exchange (ETDEWEB)

Chouraqui, A; Creuzillet, C; Barrat, J [Hopital Saint-Antoine, 75 - Paris (France)

1985-04-21

Every single person is exposed to natural (7 rads) or artificail (7.25 rads) irradiation throughout life. To which must be added, for many, irradiation from radiological examinations, which may cause malformations, genetic defects or cancer. The management of irradiated pregnant women depends on the dose received and on the age of pregnancy and requires, when the patient is seen, close co-operation between genetician, radiologist and gynaecologist. A radiological examination may be irreplaceable for diagnostic purposes, but the benefits to be expected from it should not lead to problems, particularly human problems, that are extremely difficult to solve. Non-urgent X-ray examinations should be performed outside pregnancy.

Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

International Nuclear Information System (INIS)

Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

2011-01-01

The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

Las notas al pie en la traducción de Fifty Shades (Footnotes in the Translation of Fifty Shades

Directory of Open Access Journals (Sweden)

Xinia Valverde Jara

2017-11-01

Full Text Available Se analiza el recurso de las notas a pie de página insertas en la traducción independiente de literatura comercial en versión digital, de la trilogía Fifty Shades of Grey, Fifty Shades Darker y Fifty Shades Freed, de E. L. James. Se examina el grado de influencia de esos elementos extratextuales, y se concluye que todos ellos, organizados en un complejo proceso de traducción, en especial los del contexto de llegada, condicionan la estrategia traductológica. Abstract This study examines the use of footnotes as a translation strategy in the independent translation of commercial literature in digital format, based on the trilogy written by E. L. James, Fifty Shades of Grey, Fifty Shades Darker and Fifty Shades Freed. The analysis measures the influence of these extratextual elements within the complex translation process, specifically those of the target language context, and it is concluded that they condition the translation strategies used.

Alexandre - a multi-project, multi-material and multi-technique action for an irradiation experiment in Osiris and post irradiation examination

International Nuclear Information System (INIS)

Averty, X.; Brachet, J.C.; Bertin, J.L.; Pizzanelli, J.P.; Rozenblum, F.

1999-01-01

This paper presents the data obtained on different classes of steels neutron irradiated at 325 deg C in pressurized water with a PWR-type chemistry. This irradiation, nicknamed 'Alexandre', took place in the OSIRIS reactor and finished in November 1999, for a maximum irradiation damage of ∼9 dpa. The preliminary results (up to 3.4 dpa), discussed in relation to chemical composition and initial metallurgical conditions, are listed below: - Evolution of the mechanical properties as a function of irradiation dose including the measurements of the Reduction-in-Area to failure by image analysis. - Comparison between out-of-pile and in-pile uniform corrosion. - Microstructural aspects (fractography, Transmission Electron Microscopy, and Small Angle Neutron Scattering measurements). - Post-irradiation evolution of residual. activity. (authors)

As-Run Physics Analysis for the UCSB-1 Experiment in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Nielsen, Joseph Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

The University of California Santa Barbara (UCSB) -1 experiment was irradiated in the A-10 position of the ATR. The experiment was irradiated during cycles 145A, 145B, 146A, and 146B. Capsule 6A was removed from the test train following Cycle 145A and replaced with Capsule 6B. This report documents the as-run physics analysis in support of Post-Irradiation Examination (PIE) of the test. This report documents the as-run fluence and displacements per atom (DPA) for each capsule of the experiment based on as-run operating history of the ATR. Average as-run heating rates for each capsule are also presented in this report to support the thermal analysis.

Microstructural examination of several commercial ferritic alloys irradiated to high fluence

International Nuclear Information System (INIS)

Gelles, D.S.

1981-01-01

Microstructural observations are reported for a series of five commercial ferritic alloys, 2 1/4 Cr-1 Mo, H-11, EM-12, 416, and 430F, covering the composition range 2.25 to 17% chromium, following EBR-II irradiation over the temperature range 400 to 650 0 C and to a maximum fluence of 17.6 x 10 22 n/cm 2 (E > 0.1 MeV). These materials were confirmed to be low void swelling with maximum swelling of 0.63% measured in EM-12 following irradiation at 400 0 C to 14.0 x 10 22 n/cm 2 . A wide range of precipitation response was found both as a function of alloy and irradiation temperature. Precipitates observed included M 6 C, Mo 2 C, Chi, Laves, M 23 C 6 , α' and a low temperature phase as yet unidentified. It is predicted, based on these results, that the major impact of irradiation on the ferritic alloy class will be changes in postirradiation mechanical properties due to precipitation

Microstructural examination of several commercial ferritic alloys irradiated to high fluence

Science.gov (United States)

Gelles, D. S.

Microstructural observations are reported for a series of five commercial ferritic alloys, 2 {1}/{4}Cr-1Mo , H-11, EM-12, 416, and 430F, covering the composition range 2.25 to 17% chromium, following EBR-II irradiation over the temperature range 400 to 650°C and to a maximum fluence of 1.76 × 10 23 n/cm 2 (E >0.1 MeV). These materials were confirmed to be low void swelling with maximum swelling of 0.63% measured in EM-12 following irradiation at 400°C to 1.40 × 10 23 n/cm 2. A wide range of precipitation response was found both as a function of alloy and irradiation temperature. Precipitates observed included M 6C, Mo 2C, Chi, Laves, M 23C 6, α' and a low temperature phase as yet unidentified. It is predicted, based on these results, that the major impact of irradiation on the ferritic alloy class will be changes in postirradiation mechanical properties due to precipitation.

Main results of post-irradiation examinations of new-generation fuel assemblies VVER-1000

International Nuclear Information System (INIS)

Zvir, E.; Markov, D.; Polenok, V.; Zhitelev, V.; Kobylyansky, G.

2009-01-01

To increase the competitiveness of Russian nuclear fuel at the foreign market and to improve its technical and economic performance in order to provide a necessary level of safety, it is necessary to solve certain important tasks: Increase of fuel burn-up; Extension of operational lifetime of fuel assemblies and operational reliability of nuclear fuel; Introduction of cost-beneficial and flexible fuel cycles. Alternative fuel assemblies TVSA VVER-1000 and TVS-2 are used as a basis to optimize the nuclear fuel and develop advanced fuel cycles for nuclear power plants with VVER-1000 reactor types. Four fuel assemblies TVSA operated during 1 and up to 6 reactor cycles, reference fuel assembly TVS-2 operated during three reactor cycles and achieved an average fuel burnup of 48MW·day/kgU as well as failed fuel assembly TVS-2 operated during one cycle were examined at RIAR in recent years. The main objectives of these examinations were to obtain experimental data in support of operational integrity of products or to find out reasons of their failure. The performed post-irradiation examinations confirmed the operational integrity of alternative fuel assemblies TVSA including their geometrical stability up to the average fuel burnup of 55 MW·day/kgU over the fuel assembly (FA) (up to the maximal fuel burnup of ∼73 MW·day/kgU in fuel rods) and of TVS-2 up to the average fuel burnup of 48 MW·day/kgU over the fuel assembly. The changes introduced in the design of VVER-1000 fuel assembly during the development of alternative fuel assembly TVSA and TVS-2 did not make any negative effect on fuel rods. It was proved that causes of fuel rod failure were not related to design features of fuel assemblies. The design features and operating conditions of fuel assemblies under examinations are briefly described. Post-irradiation examinations proved the geometrical stability of fuel assemblies TVSA and TVS-2 under operation up to the fuel burnup of ∼50 MW day/kgU, as for the

Post irradiation examinations on UMo full-sized plates - IRIS2 experiment

International Nuclear Information System (INIS)

Huet, F.; Noirot, J.; Marelle, V.; Dubois, S.; Boulcourt, P.; Sacristan, P.; Naury, S.; Lemoine, P.

2005-01-01

IRIS2 irradiation was the last irradiation of 4 full sized plates launched by CEA for the French UMo group to test in which operating conditions the coarse porosity forms in the UMo/Al interaction product. IRIS2 consists in four plates with high uranium loading and U-7wt%Mo atomised powder irradiated up to 60 days at OSIRIS reactor in IRIS device at a peak power of 238 W.cm -2 . The results show that in the tested conditions pillowing of the plate started from a fission density over 2.10 21 fission.cm -3 . Moreover, they show that the fission products and impurities have a key-role in the origin of the excessive plate swelling. (author)

Microstructural examination of 12% Cr martensitic stainless steel after irradiation at elevated temperatures in FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Hsu, Chen-Yih; Gelles, D.S.; Lechtenberg, T.A.

1986-06-01

A remelted 12% Cr martensitic stainless steel (HT-9) has been examined by transmission electron microscopy before and after irradiation in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The irradiation temperatures were 365,420, 520, and 600 degree C with the fluences as high as 7.3 x 10 22 n/cm 2 (E > 0.1 MeV) or 34 dpa. The extracted precipitates from each specimen were identified using x-ray microanalysis and selected area diffraction. The precipitates in the unirradiated condition were primarily M 23 C 6 carbides, which formed at martensite lath and prior austenite grain boundaries. During irradiation at elevated temperatures, small amounts of other phases formed, which were tentatively identified as the chromium-rich α', the nickel-silicon rich G-phase, and the intermetallic Chi phase. Irradiation-induced voids were observed only in specimens irradiated at 420 degree C to a dose of 34 dpa; no voids were found for specimens irradiated at 365, 520, and 600 degree C (∼11, ∼34, and ∼34 dpa). These results are not in agreement with previous experiments in that voids have not been reported in this alloy at relatively high fluence level (∼67 dpa) following irradiation in another fast-spectrum reactor (EBR.II). This is, however, the first observation following FFTF irradiation. The present results indicate that cavities can form in HT-9 at modest fluence levels even without significant generation of helium. Hence, the cavity formation in this class of ferritic alloys is not simply caused by helium generation but rather more complex mechanisms. 12 refs., 2 figs., 3 tabs

Westinghouse Advanced Doped Pellet - Characteristics and irradiation behavior

International Nuclear Information System (INIS)

Backman, K.; Hallstadius, L.; Roennberg, G.

2009-01-01

Full text: There are a number of trends in the nuclear power industry, which put additional requirements on the operational flexibility and reliability of nuclear fuel, for example power uprates and longer cycles in order to increase production, higher burnup levels in order to reduce the backend cost of the fuel cycle, and lower goals for activity release from power plant operation. These additional requirements can be addressed by increasing the fuel density, improving the FG retention, improving the PCI resistance and improving the post-failure performance. In order to achieve that, Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO 2 fuel containing additions of chromium and aluminium oxides. The additives facilitate pellet densification during sintering, enlarge the pellet grain size, and increase the creep rate. The final manufactured doped pellets reach about 0.5 % higher density within a shorter sintering time and a five times larger grain size compared with standard UO 2 fuel pellets. Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA Optima2 reloads in 2005. ADOPT pellets has been investigated in pool-side and hot cell Post Irradiation Examinations (PIEs), as well as in a ramp test and a fuel washout test in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced Fission Gas Release (FGR), improved Pellet Cladding Interaction (PCI) performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. The better FGR behaviour of ADOPT has been verified with a pool side FGR gamma measurement performed at 55 MWd/kgU, as well as transient tests in the Studsvik R2 reactor. Creep measurements performed on fresh pellets show that ADOPT has a higher creep rate which is beneficial for the PCI performance. ADOPT has also been part of a high power Halden test (IFA-677). The

HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

Energy Technology Data Exchange (ETDEWEB)

Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

1998-03-01

As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

Zīmola nozīme patērētāju lēmuma pieņemšanā mēbeļu mazumtirdzniecībā.

OpenAIRE

Pētersons, Miks

2013-01-01

Maģistra darba temats ir „Zīmola nozīme patērētāju lēmuma pieņemšanā mēbeļu mazumtirdzniecībā”. Darba mērķis ir teorētiski analizēt un praktiski noskaidrot zīmola nozīmi patērētāja lēmuma pieņemšanā par mēbeļu iegādi. Temats ir aktuāls un darbā apkopotie rezultāti var būt saistoši Latvijas mēbeļu nozares dalībniekiem. 1. nodaļā autors izvērtē patērētāju uzvedību noteicošos faktorus, kā arī zīmola nozīme lēmuma pieņemšanā. 2. nodaļā analizēta mēbeļu ražošanas nozare Latvijā, veikta uzņēm...

Effect of Farm on Productive and Reproductive Performance in Sows of Prestice Black-pied Pig

Directory of Open Access Journals (Sweden)

Pavel Nevrkla

2016-01-01

Full Text Available The aim of the study was to evaluate productive and reproductive performance of Prestice Black‑Pied sows including losses of piglets from birth to weaning in conditions of two farms. The experiment involved one hundred sows of Prestice Black‑Pied pig from the 1st to the 7th parity (50 from farm A and 50 from farm B. The evaluation of productive parameters revealed a highly statistically significant difference (P ≤ 0.001 between gilts of the observed farms in percentage of lean meat. Backfat thickness was 0.19 cm lower in gilts from the farm A, which is a very highly statistically significant difference (P ≤ 0.001. The evaluation of reproductive performance showed, that age at the time of the first insemination and farrowing was significantly higher (P ≤ 0.05 in gilts from the farm A compared to gilts from the farm B by 38 days on average. The analysis also indicates that there was no significant difference in length of gestation, total number of piglets and numbers of live-born and reared piglets between the sows of the evaluated farms. A highly statistically significant difference (P ≤ 0.01 was found in number of stillborn piglets. Interval length was significantly longer (P ≤ 0.05 in sows from the farm A, by 14.29 days. The evaluation of losses of piglets revealed a significant difference (P ≤ 0.05 between the farms in favor of piglets from the farm A.

A data recording and processing system for examination of irradiated fuel rods

International Nuclear Information System (INIS)

Olshausen, K.D.; Christiansen, J.F.; Larsen, J.K.; Stevens, T.

1978-03-01

Over the past eight years a system has been developed at Institutt for atomenergi, Kjeller, Norway, which helps to collect and analyze data obtained during certain post-irradiation examination techniques employed at IFA's Hot Laboratory. At present computer codes are in use for treating results from: profilometry, gamma scanning, eddy current testing and tensile testing. The system is easily adaptable to any othermeasuring data with the structure: -axial position, time interval etc.; signal A (DC-voltage, counts); signal B (DC-voltage, counts). Anextention of the system for transverse gamma scanning and gap measurement (mechanical compression method) is planned for the future. The various programmes in the system are described and examples of the outputs are given in a series of appendices. (JIW)

Post-irradiation examinations on the KNK II/1 fuel element NY-203 with 400 equivalent full-power days residence time and 10 % burnup

International Nuclear Information System (INIS)

Patzer, G.; Geier, F.

1984-09-01

The fuel assembly NY-203 has been irradiated in the first core of KNK II up to a burnup of about 10 % and a residence time of 400 equivalent full-power days. The assembly contained 211 fuel pins with 6.0 mm outer diameter and fuel pellets with the composition (U 0 .7Pu 0 .3)O 2 .00. The cladding material was the austenitic steel 1.4988 lg. Some selected pins were examined in the hot cells of the KfK Karlsruhe. The post-irradiation examinations did not reveal any critical design aspects [de

The effect of zinc addition on PWR corrosion product deposition on zircaloy-4

International Nuclear Information System (INIS)

Walters, W.S.; Page, J.D.; Gaffka, A.P.; Kingsbury, A.F.; Foster, J.; Anderson, A.; Wickenden, D.; Henshaw, J.; Zmitko, M.; Masarik, V.; Svarc, V.

2002-01-01

During the period 1995 to 2001 a programme of loop irradiation tests have been performed to confirm the effectiveness of zinc additions on PWR circuit chemistry and corrosion. The programme included two loop irradiation experiments, and subsequent PIE; the experiments were a baseline test (no added zinc) and a test with added zinc (10 ppb). This paper addresses the findings regarding corrosion product deposition and activation on irradiated Zircaloy-4 surfaces. The findings are relevant to overall corrosion of the reactor primary circuit, the use of zinc as a corrosion inhibitor, and activation and transport of corrosion products. The irradiation experience provides information on the equilibration of the loop chemistry, with deliberate injection of zinc. The PIE used novel and innovative techniques (described below) to obtain samples of the oxide from the irradiated Zircaloy. The results of the PIE, under normal chemistry and zinc chemistry, indicate the effect of zinc on the deposition and activation of corrosion products on Zircaloy. It was found that corrosion product deposition on Zircaloy is enhanced by the addition of zinc (but corrosion product deposition on other materials was reduced in the presence of zinc). Chemical analysis and radioisotope gamma counting results are presented, to interpret the findings. A computer model has also been used to simulate the corrosion product deposition and activation, to assist in the interpretation of the results. (authors)

Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

International Nuclear Information System (INIS)

Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

1999-06-01

Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

Gamma spectrometrical examination of irradiated fuel

International Nuclear Information System (INIS)

Kristof, Edvard; Pregl, Gvido

1988-01-01

Gamma scanning is the only non-destructive technique for quantitative measuring of fission or activation products in spent fuel. The negligence of local variation of the linear attenuation coefficient of gamma rays in the irradiated fuel remains the main source of systematic error. To eliminate it we combine the (single) emission gamma ray scanning technique with a transmission measurement. Mathematical procedure joined with the experiment is particularly convenient for fuel elements of circular cross-section. In such a manner good results are obtainable even for relatively small number of measuring data. Accomplished routines enable to esteem the finite width of the collimation slit. The experiment has been partially automated. Trial measurements were carried out, and the measured data were successfully processed

Fabrication and testing of uranium nitride fuel for space power reactors

Science.gov (United States)

Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

1988-02-01

Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

Postirradiation examination of a low enriched U3Si2-Al fuel element manufactured and irradiated at Batan, Indonesia

International Nuclear Information System (INIS)

Suripto, A.; Sugondo, S.; Nasution, H.

1994-01-01

The first low-enriched U 3 Si 2 -Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235 U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U 3 Si 2 -Al plate-type fuel produced and tested at other institutions. The main effect of irradiation on the fuel plates is a thickness increase of 30--40 μm (2.5-3.0%). This thickness increase is almost entirely due to the formation of a corrosion layer (Boehmite). The contribution of fuel swelling to the thickness increase is rather small (less than 10 μm) commensurate with the burnup of the fuel and the relatively moderate as-fabricated fuel volume fraction of 27% in the fuel meat

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)

Fractographic examination of reduced activation ferritic/martensitic steel charpy specimens irradiated to 30 dpa at 370{degrees}C

Energy Technology Data Exchange (ETDEWEB)

Gelles, D.S.; Hamilton, M.L. [Pacific Northwest National Lab., Richland, WA (United States); Schubert, L.E. [Univ. of Missouri, Rolla, MO (United States)

1996-10-01

Fractographic examinations are reported for a series of reduced activation ferritic/Martensitic steel Charpy impact specimens tested following irradiation to 30 dpa at 370{degrees}C in FFTF. One-third size specimens of six low activation steels developed for potential application as structural materials in fusion reactors were examined. A shift in brittle fracture appearance from cleavage to grain boundary failure was noted with increasing manganese content. The results are interpreted in light of transmutation induced composition changes in a fusion environment.

An experiment to examine the mechanistic behaviour of irradiated CANDU fuel stored under dry conditions

International Nuclear Information System (INIS)

Oldaker, I.E.; Crosthwaite, J.L.; Keltie, R.J.; Truss, K.J.

1979-01-01

A program has begun to use the Whiteshell Nuclear Research Establishment dry-storage canisters to store some selected CANDU irradiated fuel bundles in an 'easily retrievable basket.' The object of the experimental program is to study the long-term stability of the Zircaloy-sheathed UO 2 and UC fuel elements when stored in air. Bundles were loaded into a canister in October 1979 following detailed examination and removal of up to three complete elements from most bundles. These elements are currently being subjected to detailed destructive examinations, including metallography and scanning electron micrography, to fully characterize their pre-storage condition. After four years, and every five years thereafter, further elements will be examined similarly to study the effects of the storage environment on the stability of the Zircaloy sheathing, and on its continued ability to contain the fuel safely in an interim storage facility. (author)

Preliminary examination of induced radio activity in pepper by 10 MeV electron irradiation

International Nuclear Information System (INIS)

Furuta, Masakazu; Katayama, Tadashi; Ito, Norio; Mizohata, Akira; Matsunami, Tadao; Toratani, Hirokazu; Takeda, Atsuhiko

1989-01-01

β-ray measurement was performed on 10 MeV electron-irradiated black pepper and white pepper in order to reconfirm the wholesomeness of irradiated food and present unambiguous data to general consumers concerning about the induced radioactivity in the irradiated foods. From elemental composition of the samples and investigation of photonuclear reactions, several β-emmitters were listed up. But no radioactivity other than from natural sources was detected in the irradiated sample by β-ray counting with 2 π gass flow counter, suggesting that the induced β-emmitters in the irradiated sample was below the detection limit of its induced radioactivity. (author)

Examination into the gamma irradiation of activated sludge

International Nuclear Information System (INIS)

Mustapha, S.; Forster, C.F.

1985-01-01

This study has shown that the treatment of activated sludge by gamma irradiation resulted in a deterioration in the filterability, a decrease in the size of the floc particles and an increase in the organic matter present in the sludge supernatant. A significant difference was found between the results obtained for filamentous and non-filamentous sludges in relation to the amount of soluble polysaccharide produced. (author)

Uz nestriktas loģikas balstīta lēmumu pieņemšanas sistēma biznesa spēlēs

OpenAIRE

Vitkovska, Diāna

2015-01-01

Darbs ir veltīts uz nestriktās loģikas balstītas lēmumu pieľemšanas sistēmas pielietojumam stratēģiju izvēlei biznesa spēlēs. Darbā tika aplūkotas divu personu spēles kā matemātiskais modelis divu uzľēmumu konkurences analīzei. Tika aprakstīta stratēģiju izvēles metode, kas balstās uz nestriktās secinājumu sistēmas, izmanto sistēmas nestrikto ieejas un izejas mainīgo lingvistiskās vērtības un lēmumu pieľemšanas kārtulas un tiek īstenota programmatūras Matlab vidē. Šādas metodes efektivitāte i...

Ana?lisis cinemático de la marcha en pacientes con pie zambo tratados mediante el me?todo de Ponseti frente a la te?cnica quiru?rgica de liberacio?n posterior

OpenAIRE

Ferrando, A.; Salom Taverner, M.; Page, A.

2018-01-01

El objetivo principal de este proyecto consiste en valorar la evolución de la marcha en niños en edad preadolescente tratados mediante el método de Ponseti frente a los tratados mediante liberación posterior a partir de técnicas de valoración de la marcha mediante análisis biomecánico. Material y Métodos Estudio retrospectivo de casos y controles aprobado por el comité de ética. Grupo 1: 28 niños (39 pies) tratados mediante liberación posterior. Grupo 2: 18 pacientes (31 pies) tratados median...

Lead exposure and biological effects in pied flycatchers (Ficedula hypoleuca) before and after the closure of a lead mine in northern Sweden

International Nuclear Information System (INIS)

Berglund, A.M.M.; Ingvarsson, P.K.; Danielsson, H.; Nyholm, N.E.I.

2010-01-01

Mining activities affect the surrounding environment by increasing exposure to metals. In this study, metal accumulation and its effects on reproduction and health of pied flycatcher (Ficedula hypoleuca) nestlings were monitored before and up to five years after a lead mine and enrichment plant closed down. The lead concentration in moss, nestling blood, liver and feces all indicated decreased lead exposure by at least 31% after closure, although only blood lead decreased significantly. Although the birds responded fairly well to the changed atmospheric deposition (based on moss samples), concentrations were still higher compared with birds in a reference area, and breeding was affected at the mine (smaller clutches and higher mortality). Surviving nestlings suffered from lower hemoglobin levels, mean cell hemoglobin concentrations and inhibited δ-aminolevulinic acid dehydratase activity. Lead poisoning contributed to poor health and adverse reproductive effects, but other factors (e.g. increased parasitic load) probably also affected the birds. - Increased mortality and lower blood values in pied flycatcher (Ficedula hypoleuca) nestlings near a closed lead mine.

Full-sized plates irradiation with high UMo fuel loading. Final results of IRIS 1 experiment

International Nuclear Information System (INIS)

Huet, F.; Marelle, V.; Noirot, J.; Sacristan, P.; Lemoine, P.

2003-01-01

As a part of the French UMo Group qualification program, IRIS 1 experiment contained full-sized plates with high uranium loading in the meat of 8 g.cm -3 . The fuel particles consisted of 7 and 9 wt% Mo-uranium alloys ground powders. The plate were irradiated at OSIRIS reactor in IRIS device up to 67.5% peak burnup within the range of 136 W.cm - '2 for the heat flux and 72 deg. C for the cladding temperature. After each reactor cycle the plates thickness were measured. The results show no swelling behaviour differences versus burnup between UMo7 and UMo9 plates. The maximum plate swelling for peak burnup location remains lower than 6%. The wide set of PIE has shown that, within the studied irradiation conditions, the interaction product have a global formulation of '(U-Mo)Al -7 ' and that there is no aluminium dissolution in UMo particles. IRIS1 experiment, as the first step of the UMo fuel qualification for research reactor, has established the good behaviour of UMo7 and UMo9 high uranium loading full-sized plate within the tested conditions. (author)

Irradiation and examination results of the AC-3 mixed-carbide test

International Nuclear Information System (INIS)

Mason, R.E.; Hoth, C.W.; Stratton, R.W.; Botta, F.

1992-01-01

The AC-3 test was a cooperative Swiss/US irradiation test of mixed-carbide, (U,Pr)C, fuel pins in the Fast Flux Test Facility. The test included 25 Swiss-fabricated sphere-pac-type fuel pins and 66 U.S. fabricated pellet-type fuel pins. The test was designed to operate at prototypical fast reactor conditions to provide a direct comparison of the irradiation performance of the two fuel types. The test design and fuel fabrication processes used for the AC-3 test are presented