78 FR 13662 - Cobra Pipeline Ltd.; Notice of Petition

Science.gov (United States)

2013-02-28

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket Nos. PR13-32-000; PR13-33-000] Cobra Pipeline Ltd.; Notice of Petition Take notice that on February 4, 2013, Cobra Pipeline Ltd. (Cobra... Docket No. PR13-11-000, as more fully detailed in the petitions. Any person desiring to participate in...

Popsicle-Stick Cobra Wave.

Science.gov (United States)

Boucher, Jean-Philippe; Clanet, Christophe; Quéré, David; Chevy, Frédéric

2017-08-25

The cobra wave is a popular physical phenomenon arising from the explosion of a metastable grillage made of popsicle sticks. The sticks are expelled from the mesh by releasing the elastic energy stored during the weaving of the structure. Here we analyze both experimentally and theoretically the propagation of the wave front depending on the properties of the sticks and the pattern of the mesh. We show that its velocity and its shape are directly related to the recoil imparted to the structure by the expelled sticks. Finally, we show that the cobra wave can only exist for a narrow range of parameters constrained by gravity and rupture of the sticks.

COBRA-IV wire wrap data comparisons

International Nuclear Information System (INIS)

Donovan, T.E.; George, T.L.; Wheeler, C.L.

1979-02-01

Thermal hydraulic analyses of hexagonally packed wire-wrapped fuel assemblies are complicated by the induced crossflow between adjacent subchannels. The COBRA-IV computer code simultaneously solves the hydrodynamics and thermodynamics of fuel assemblies. The modifications and the results are presented which are predicted by the COBRA-IV calculation. Comparisons are made with data measured in five experimental models of a wire-wrapped fuel assembly

Steady state thermal hydraulic analysis of LMR core using COBRA-K code

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

1997-02-01

A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

Energy Technology Data Exchange (ETDEWEB)

Wheeler, C.L.; Stewart, C.W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

1976-03-01

The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels.

Cobra Fiber-Optic Positioner Upgrade

Science.gov (United States)

Fisher, Charles D.; Braun, David F.; Kaluzny, Joel V.

2013-01-01

A prime focus spectrometer (PFS), along with corrective optics, will mount in place of the secondary mirror of the Subaru telescope on Mauna Kea, Hawaii. This will allow simultaneous observations of cosmologic targets. It will enable large-scale galactic archeology and dark energy surveys to help unlock the secrets of the universe. To perform these cosmologic surveys, an array of 2,400 optical fibers needs to be independently positioned within the 498-mm-diameter focal plane of the PFS instrument to collect light from galaxies and stars for spectrographic analyses. To allow for independent re-positioning of the fibers, a very small positioner (7.7 mm in diameter) is required. One hundred percent coverage of the focal plane is also required, so these small actuators need to cover a patrol region of 9.5 mm in diameter. To optimize the amount of light that can be collected, the fibers need to be placed within 5 micrometers of their intended target (either a star or galaxy). The Cobra Fiber Positioner was designed to meet the size and accuracy requirements stated above. Cobra is a two-degrees-of-freedom mechanism that can position an optical fiber in the focal plane of the PFS instrument to a precision of 5 micrometers. It is a theta-phi style positioner containing two rotary piezo tube motors with one offset from the other, which enables the optic fibers to be placed anywhere in a small circular patrol region. The patrol region of the actuator is such that the array of 2,400 positioners allows for full coverage of the instrument focal plane by overlapping the patrol areas. A second-generation Cobra positioner was designed based on lessons learned from the original prototype built in 2009. Improvements were made to the precision of the ceramic motor parts, and hard stops were redesigned to minimize friction and prevent jamming. These changes resulted in reducing the number of move iterations required to position the optical fiber within 5 micrometers of its target. At

Comparison of MATRA-S and COBRA-SFS for Low Flow Subchannel Analysis

Energy Technology Data Exchange (ETDEWEB)

Seo, Kyong Won; Kwon, Hyuk; Kim, Seong Jin; Hwang, Dae Hyun [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, we compared the MATRA-S with COBRA-SFS for the PNL test because the COBRASFS is believed to be superior to MATRA-S for the low flow conditions. COBRA-SFS code was developed for subchannel analysis of spent fuel storage system based on COBRA-3C, COBRA-4I, and COBRA-WC. As the code was designed to predict temperature and flow distributions in spent fuel storage system, it can analyze thermal hydraulic fields of natural convection as well as radiation and conduction heat transfer. In the way of improving XSHCME of MATRA-S to be applicable to low flow problems, we compared MATRA-S XSCHEM and COBRA-SFS RECIRC for steady state and flow transient. Both methods use similar algorithms to solve pressure, axial flow and cross flow. MATRA-S XSCHEM predicted flow velocity profile well even negative flow in recirculation flow.

A study on the numerical instability of COBRA-series subchannel analysis codes at low-pressure and low-flow conditions

International Nuclear Information System (INIS)

Yoo, Y. J.; Hwnag, T. H.; Kim, K. K.; Ji, S. K.

2001-01-01

The numerical instability at low-pressure and low-flow conditions has been confirmed to be the common problem of the existing COBRA-series subchannel analysis codes. In addition, the range of operating conditions at which the analyses by the codes are impossible has been evaluated. To evaluate the MATRA's inapplicable range of operating conditions of the SMART core that is to be operated at the low flow condition, i.e. about 30% of the flow of the existing commercial pressurized water reactors at the steady-state condition, the analyses of various operating conditions were performed by using several representative COBRA-series subchannel analysis codes including MATRA. TORC of CE, COBRA3CP of Siemens/KWU, COBRA4I of PNL, and MATRA of KAERI were chosen as the subchannel analysis codes to be evaluated. The various operating conditions used in the CHF tests carried out at the Winfrith Establishment of UKAEA were chosen as the conditions to be analyzed. As the result, the numerical instabilities at low-pressure and low-flow conditions occurred in the analyses by all of the codes. It was revealed that the MATRA code, which numerically more stable thatn the other codes, was not able to analyze the conditions of the pressure not more than 100 bar and the mass velocity not more than 300 kg/sec-m 2 . Hereafter it is required to find out the exact reason for the numerical instability of the existing COBRA-series subchannel analysis codes at low-pressure and low-flow conditions and to devise the new method to get over that numerical problem

Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

International Nuclear Information System (INIS)

Daverio, Hernando J.

2003-01-01

Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

Verification and validation of COBRA-SFS transient analysis capability

International Nuclear Information System (INIS)

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

1998-05-01

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

Thermo-hydraulic simulations of the experimental fast reactor core

International Nuclear Information System (INIS)

Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

1985-01-01

A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

26 CFR 54.4980B-5 - COBRA continuation coverage.

Science.gov (United States)

2010-04-01

... employee and spouse who have no children divorce on May 1, 2001, and the spouse elects COBRA continuation... divorce on June 1, 2001, and one of the children remains with the employee. The spouse elects COBRA... other reference in §§ 54.4980B-1 through 54.4980B-10 to coverage in effect immediately before (or on the...

AN/FPS-108 COBRA DANE Space Surveillance Mission Evolution

Science.gov (United States)

Chorman, P.; Boggs, J.

2013-09-01

It has been ten years since the COBRA DANE radar was restored to continuous full power operations in a more dedicated role of space debris tracking. Over this time, the satellite catalog population has grown and the overall average RCS value of cataloged objects has decreased dramatically, due to a combination of breakups and collisions together with the increased sensitivity offered by COBRA DANE's support to the network. This shift in catalog composition places new challenges on COBRA DANE and other debris tracking radars (PARCS and Eglin/FPS-85) to consistently track the ever-increasing number of small objects. Space Surveillance Network radars now operate at the limits of their detection performance, tracking several thousand new objects in a size category that only the most powerful and sensitive radars can observe (i.e., COBRA DANE's inherent Spacetrack mission software functionality remained better tuned for its original support role against the larger (known) orbital objects than for its more modern role in acquiring and reporting small debris in an appreciable number -- that is, until now. Several newly-identified software changes offer promise of significantly increased data yield that will make COBRA DANE an even more important asset for this evolving mission. In the course of assisting JSpOC, AFSPC, and USSTRATCOM with the ongoing challenges of lost satellite management, it was discovered that the radar's performance is being artificially restricted by mission software, rather than by the system's overall architectural design (power-aperture envelope and radar resources). This paper captures specific opportunities to improve COBRA DANE's Spacetrack mission performance, several of which are currently implemented and slated to become operational with the next two software releases. With one of the more prominent enhancements, COBRA DANE will be capable of autonomously 'fence tasking' all newly acquired small objects. Under the current operating paradigm

Detecting surface events at the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Tebruegge, Jan [Exp. Physik IV, TU Dortmund (Germany); Collaboration: COBRA-Collaboration

2015-07-01

The aim of the COBRA experiment is to prove the existence of neutrinoless double-beta-decay and to measure its half-life. For this purpose the COBRA demonstrator, a prototype for a large-scale experiment, is operated at the Gran Sasso Underground Laboratory (LNGS) in Italy. The demonstrator is a detector array made of 64 Cadmium-Zinc-Telluride (CdZnTe) semiconductor detectors in the coplanar grid anode configuration. Each detector is 1**1 ccm in size. This setup is used to investigate the experimental issues of operating CdZnTe detectors in low background mode and identify potential background components. As the ''detector=source'' principle is used, the neutrinoless double beta decay COBRA searches for happens within the whole detector volume. Consequently, events on the surface of the detectors are considered as background. These surface events are a main background component, stemming mainly from the natural radioactivity, especially radon. This talk explains to what extent surface events occur and shows how these are recognized and vetoed in the analysis using pulse shape discrimination algorithms.

Proteomic characterization of venom of the medically important Southeast Asian Naja sumatrana (Equatorial spitting cobra).

Science.gov (United States)

Yap, Michelle Khai Khun; Fung, Shin Yee; Tan, Kae Yi; Tan, Nget Hong

2014-05-01

The proteome of Naja sumatrana (Equatorial spitting cobra) venom was investigated by shotgun analysis and a combination of ion-exchange chromatography and reverse phase HPLC. Shotgun analysis revealed the presence of 39 proteins in the venom while the chromatographic approach identified 37 venom proteins. The results indicated that, like other Asiatic cobra venoms, N. sumatrana contains large number of three finger toxins and phospholipases A2, which together constitute 92.1% by weight of venom protein. However, only eight of the toxins can be considered as major venom toxins. These include two phospholipases A2, three neurotoxins (two long neurotoxins and a short neurotoxin) and three cardiotoxins. The eight major toxins have relative abundance of 1.6-27.2% venom proteins and together account for 89.8% (by weight) of total venom protein. Other venom proteins identified include Zn-metalloproteinase-disintegrin, Thaicobrin, CRISP, natriuretic peptide, complement depleting factors, cobra venom factors, venom nerve growth factor and cobra serum albumin. The proteome of N. sumatrana venom is similar to proteome of other Asiatic cobra venoms but differs from that of African spitting cobra venom. Our results confirm that the main toxic action of N. sumatrana venom is neurotoxic but the large amount of cardiotoxins and phospholipases A2 are likely to contribute significantly to the overall pathophysiological action of the venom. The differences in toxin distribution between N. sumatrana venom and African spitting cobra venoms suggest possible differences in the pathophysiological actions of N. sumatrana venom and the African spitting cobra venoms, and explain why antivenom raised against Asiatic cobra venom is not effective against African spitting cobra venoms. Copyright © 2014 Elsevier B.V. All rights reserved.

Why are Dutch rheumatologists reluctant to use the COBRA treatment strategy in early rheumatoid arthritis?

Science.gov (United States)

van Tuyl, Lilian H D; Plass, Anne Marie C; Lems, Willem F; Voskuyl, Alexandre E; Dijkmans, Ben A C; Boers, Maarten

2007-01-01

Background The Combinatietherapie Bij Reumatoide Artritis (COBRA) trial has proved that combination therapy with prednisolone, methotrexate and sulphasalazine is superior to sulphasalazine monotherapy in suppressing disease activity and radiological progression of early rheumatoid arthritis (RA). In addition, 5 years of follow‐up proved that COBRA therapy results in sustained reduction of the rate of radiological progression. Despite this evidence, Dutch rheumatologists seem reluctant to prescribe COBRA therapy. Objective To explore the reasons for the reluctance in Dutch rheumatologists to prescribe COBRA therapy. Methods A short structured questionnaire based on social–psychological theories of behaviour was sent to all Dutch rheumatologists (n = 230). Results The response rate was 50%. COBRA therapy was perceived as both effective and safe, but complex to administer. Furthermore, rheumatologists expressed their concern about the large number of pills that had to be taken, the side effects of high‐dose prednisolone and the low dose of methotrexate. Although the average attitude towards the COBRA therapy was slightly positive (above the neutral point), the majority of responding rheumatologists had a negative intention (below the neutral point) to prescribe COBRA therapy in the near future. Conclusion The reluctance of Dutch rheumatologists to prescribe effective COBRA therapy may be due to perceptions of complexity of the treatment schedule and negative patient‐related consequences of the therapy. PMID:17392349

Performance Evaluation of the COBRA GEM for the Application of the TPC

Science.gov (United States)

Terasaki, Kohei; Hamagaki, Hideki; Gunji, Taku; Yamaguchi, Yorito

2014-09-01

Suppression of the back-drifting ions from avalanche region to drift space (IBF: Ion Backflow) is the key for a Time Projection Chamber (TPC) since IBF easily distorts the drift field. To suppress IBF, Gating Grid system is widely used for the TPC but this limits the data taking rate. Gas Electron Multiplier (GEM) has advantages in the reduction of IBF and high rate capability. By adopting GEM, it is possible to run a TPC continuously under high rate and high multiplicity conditions. Motivated by the study of IBF reduction for RICH with Thick COBRA, which has been developed by F. A. Amero et al., we developed COBRA GEMs for the application of a TPC. With a stack configuration, IBF reaches about 0.1 ~ 0.5%, which is ×5--10 better IBF than the standard GEMs. However, the measured energy resolution with COBRA is 20% (σ) and this is much worse than the resolution with standard GEMs. Measurement of long-time stability of gain indicates that gain of COBRA varies significantly due to charging up effect. Simulation studies based on Garfield++ are performed for understanding quantitatively the reasons of worse energy resolution and instability of gain. In this presentation, we will report the simulation studies together with the measured performance of the COBRA GEM.

Cobra-Gruppens besøg i Bregnerød

DEFF Research Database (Denmark)

Baumeister, Ruth

2014-01-01

This article describes the Cobra-meeting in Bregnerød. It exemplifies the interior painting of the weekend retreat of the Royal academy´s architecture students and contextualizes the event within the European postwar discourse of a synthesis of the arts.......This article describes the Cobra-meeting in Bregnerød. It exemplifies the interior painting of the weekend retreat of the Royal academy´s architecture students and contextualizes the event within the European postwar discourse of a synthesis of the arts....

Coiled Brine Recovery Assembly (CoBRA) Project

Data.gov (United States)

National Aeronautics and Space Administration — The Coiled Brine Recovery Assembly (CoBRA) project will result in a proof-of-concept demonstration for a lightweight, compact, affordable, regenerable and disposable...

Discovery Of Human Antibodies Against Spitting Cobra Toxins

DEFF Research Database (Denmark)

Bojsen-Møller, Laura; Lohse, Brian; Harrison, Robert

Current snakebite envenoming treatment options consist of animal-derived antisera and are associated with severe adverse reactions due to the heterologous nature of the animal-derived antibodies present in these antisera, and the presence of therapeutically irrelevant antibodies. The African...... spitting cobras are among the most medically important snakes in sub-Saharan regions due to the severity of the clinical outcomes caused by their cytotoxic venom, which is derived from cytotoxins of the 3FTx toxin family and PLA2. Here we report the results of our progress in identifying human antibodies...... targeting relevant toxins from the venom of the black necked spitting cobra (Naja nigricolis)....

Discordant perspectives of rheumatologists and patients on COBRA combination therapy in rheumatoid arthritis.

NARCIS (Netherlands)

Tuyl, L.H.D. van; Plass, A.M.C.; Lems, W.F.; Voskuyl, A.E.; Kerstens, P.J.S.M.; Dijkmans, B.A.C.; Boers, M.

2008-01-01

Objective. The COBRA therapy (combination therapy in early rheumatoid arthritis) has proven to be an effective treatment for early RA, but is rarely prescribed. A survey showed reluctance of Dutch reumatologists to apply COBRA therapy in early RA. The present qualitative study was carried out to

Improvements, enhancements, and optimizations of COBRA-TF

International Nuclear Information System (INIS)

Salko, R. K.; Avramova, M. N.; Hooper, R.; Palmtag, S.; Popov, E.; Turner, J.

2013-01-01

The Reactor Dynamics and Fuel Management Group (RDFMG) at The Pennsylvania State University (PSU) has become active in the Consortium for Advanced Simulation of Light Water Reactors (CASL) program by delivering, supporting, and further developing CTF, the PSU version of the Coolant Boiling in Rod Arrays - Two Fluids (COBRA-TF) Thermal/Hydraulic (T/H), sub-channel program. New development work on CTF was primarily geared towards decreasing the execution time of the code so that it may eventually be used for performing pin-by-pin, full-core simulations. Great gains have been made through targeting sections of source code for optimization. For example, wall clock time has been reduced for a one-assembly, three-dimensional model, containing ∼9,400 mesh cells, from 9.2 min to 1 min. A further improvement has been reduction in code memory usage, which was previously prohibitive for large models. In conjunction with the run time speedups, this has enabled the simulation of a refined quarter-core model (∼460,000 mesh cells), which saw a reduction in memory usage from over 130 GB to less than 3 GB. In addition to the optimization work, RDFMG has also created a preprocessor utility for the fast and less error-prone generation of CTF input decks. Furthermore, basic post-processing capabilities have been implemented by creating a CTF subroutine for producing Visualization Tool-Kit (VTK) files that output mesh data and associated simulation results. These VTK files can be opened with visualization software. (authors)

Thermohydraulic analysis of pressurized water reactors

International Nuclear Information System (INIS)

Veloso, M.A.

1980-01-01

The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

Synchronization and an application of a novel fractional order King Cobra chaotic system

Energy Technology Data Exchange (ETDEWEB)

Muthukumar, P., E-mail: muthukumardgl@gmail.com; Balasubramaniam, P., E-mail: balugru@gmail.com [Department of Mathematics, Gandhigram Rural Institute‐Deemed University, Gandhigram 624 302, Tamilnadu (India); Ratnavelu, K., E-mail: kuru052001@gmail.com [Faculty of Science, Institute of Mathematical Sciences, University of Malaya, 50603 Kuala Lumpur (Malaysia)

2014-09-01

In this paper, we design a new three dimensional King Cobra face shaped fractional order chaotic system. The multi-scale synchronization scheme of two fractional order chaotic systems is described. The necessary conditions for the multi-scale synchronization of two identical fractional order King Cobra chaotic systems are derived through feedback control. A new cryptosystem is proposed for an image encryption and decryption by using synchronized fractional order King Cobra chaotic systems with the supports of multiple cryptographic assumptions. The security of the proposed cryptosystem is analyzed by the well known algebraic attacks. Numerical simulations are given to show the effectiveness of the proposed theoretical results.

MagLev Cobra: Test Facilities and Operational Experiments

Science.gov (United States)

Sotelo, G. G.; Dias, D. H. J. N.; de Oliveira, R. A. H.; Ferreira, A. C.; De Andrade, R., Jr.; Stephan, R. M.

2014-05-01

The superconducting MagLev technology for transportation systems is becoming mature due to the research and developing effort of recent years. The Brazilian project, named MagLev-Cobra, started in 1998. It has the goal of developing a superconducting levitation vehicle for urban areas. The adopted levitation technology is based on the diamagnetic and the flux pinning properties of YBa2Cu3O7-δ (YBCO) bulk blocks in the interaction with Nd-Fe-B permanent magnets. A laboratory test facility with permanent magnet guideway, linear induction motor and one vehicle module is been built to investigate its operation. The MagLev-Cobra project state of the art is presented in the present paper, describing some construction details of the new test line with 200 m.

Plastic surgical management of a cobra bite – a case study

Directory of Open Access Journals (Sweden)

Kuhbier, Jörn W.

2017-02-01

Full Text Available Cobra bites are quite rare in European countries as these snakes are not native there. Toxins are devastating for tissue resulting in massive necrosis, thus plastic surgery might play a role in reconstruction of the lost tissue. A case of a male patient bitten by a thai cobra in the left index finger is presented. Antitoxin administration was delayed due to secondary patient admission. Progressive tissue necrosis made radical debridement necessary, resulting in the need for plastic surgical defect coverage with a flap. While a radical debridement to prevent toxic necrosis due to lytic enzymes in cobra venom has been favoured beforehand, large case studies led to a more restrained initial surgical intervention. However, antitoxin administration should be first line therapy in management of these cases. If severe necrosis is present as it might occur in delayed admission, a plastic surgical management of the patient might be advantageous.

Irradiated cobra (Naja naja) venom for biomedical applications

International Nuclear Information System (INIS)

Kankonkar, S.R.; Kankonkar, R.C.; Gaitonde, B.B.

1975-01-01

Ionizing radiation is known to cause damage to proteins in aqueous solutions in a selective manner, thereby producing remarkable changes in their properties. Since venoms are very rich in proteins, it was felt that they would also show such changes upon irradiation. It was of interest to know if one could get rid of the toxicity and retain the immunogenicity of the venom by suitable choice of radiation dose and strength of venom solution. If so, the method could be profitably exploited for the rapid preparation of venom toxoid and this could be expected to have many applications in the biological sciences. Accordingly, laboratory investigations were undertaken on the effect of gamma radiation on cobra (Naja naja) venom. To avoid drastic changes, solutions of cobra venom having low protein content were irradiated with gamma radiation from a cobalt-60 source. The results obtained with 0.01 to 1.0% venom solutions are found to be encouraging. The solutions did not manifest any toxicity in mice. For the immunogenicity test, guinea pigs were immunized with varying doses of the irradiated cobra venom and the immunized guinea pigs were found to survive when challenged with as big a dose as 10 MLD (i.e. minimum lethal dose, approximately 1 mg). The paper describes the experimental details and the results of the observations. (author)

Assessment of reflood heat transfer model of COBRA-TIF against ABB-CE evaluation model

Energy Technology Data Exchange (ETDEWEB)

Lee, S. I.; Lee, S. Y.; Park, C. E.; Choi, H. R.; Choi, C. J. [Korea Power Engineering Company Inc., Taejon (Korea, Republic of)

2000-05-01

According to 10 CFR 50 Appendix K, ECCS performance evaluation model should be based on the experimental data of FLECHT and have the conservatism compared with experimental data. To meet this requirement ABB-CE has the complicate code structure as follows: COMPERC-II calculates three reflood rates, and FLELAPC and HTCOF calculate the reflood heat transfer coefficients, and finally STRIKIN-II calculates the cladding temperature using the reflood heat transfer calculated in previous stage. In this paper, to investigate whether or not COBRA-TF satisfies the requirement of Appendix K, the reflood heat transfer coefficient of COBRA-TF was assessed against ABB-CE MOD-2C model. It was found out that COBRA-TF predicts properly the experimental data and has more conservatism than the results of ABB-CE MOD-2C model. Based on these results, it can be concluded that the reflood heat transfer coefficients calculated by COBRA-TF meet the requirement of Appendix K.

MagLev Cobra: Test Facilities and Operational Experiments

International Nuclear Information System (INIS)

Sotelo, G G; Dias, D H J N; De Oliveira, R A H; Ferreira, A C; De Andrade, R Jr; Stephan, R M

2014-01-01

The superconducting MagLev technology for transportation systems is becoming mature due to the research and developing effort of recent years. The Brazilian project, named MagLev-Cobra, started in 1998. It has the goal of developing a superconducting levitation vehicle for urban areas. The adopted levitation technology is based on the diamagnetic and the flux pinning properties of YBa 2 Cu 3 O 7−δ (YBCO) bulk blocks in the interaction with Nd-Fe-B permanent magnets. A laboratory test facility with permanent magnet guideway, linear induction motor and one vehicle module is been built to investigate its operation. The MagLev-Cobra project state of the art is presented in the present paper, describing some construction details of the new test line with 200 m.

Benchmarking of LOFT LRTS-COBRA-FRAP safety analysis model

International Nuclear Information System (INIS)

Hanson, G.H.; Atkinson, S.A.; Wadkins, R.P.

1982-05-01

The purpose of this work was to check out the LOFT LRTS/COBRA-IV/FRAP-T5 safety-analysis models against test data obtained during a LOFT operational transient in which there was a power and fuel-temperature rise. LOFT Experiment L6-3 was an excessive-load-increase anticipated transient test in which the main steam-flow-control valve was driven from its operational position to full-open in seven seconds. The resulting cooldown and reactivity-increase transients provide a good benchmark for the reactivity-and-power-prediction capability of the LRTS calculations, and for the fuel-bundle and fuel-rod temperature-response analysis capability of the LOFT COBRA-IV and FRAP-T5 models

Cobra: A content-based video retrieval system

NARCIS (Netherlands)

Petkovic, M.; Jonker, W.; Jensen, C.S.; Jeffery, K.G.; Pokorny, J.; Saltenis, S.; Bertino, E.; Böhm, K.; Jarke, M.

2002-01-01

An increasing number of large publicly available video libraries results in a demand for techniques that can manipulate the video data based on content. In this paper, we present a content-based video retrieval system called Cobra. The system supports automatic extraction and retrieval of high-level

Conversion of the COBRA-IV-I code from CDC CYBER to HP 9000/700 version

International Nuclear Information System (INIS)

Sohn, D. S.; Yoo, Y. J.; Nahm, K. Y.; Hwang, D. H.

1996-01-01

COBRA-IV-I is a multichannel analysis code for the thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores based on the subchannel approach. The existing COBRA-IV-I code is the control data corporation (CDC) CYBER version, which has limitations on the computer core storage and gives some inconvenience to the user interface. To solve these problems, we have converted the COBRA-IV-I code form the CDC CYBER mainframe to an Hewlett Packard (HP) 9000/700-series workstation version, and have verified the converted code. as a result, we have found almost no difference between the two versions in their calculation results. Therefore we expect the HP 9000/700 version of the COBRA-IV-I code to be the basis for the future development of an improved multichannel analysis code under the more convenient user environment. (author). 3 tabs., 2 figs., 8 refs

COBRA: a Bayesian approach to pulsar searching

Science.gov (United States)

Lentati, L.; Champion, D. J.; Kramer, M.; Barr, E.; Torne, P.

2018-02-01

We introduce COBRA, a GPU-accelerated Bayesian analysis package for performing pulsar searching, that uses candidates from traditional search techniques to set the prior used for the periodicity of the source, and performs a blind search in all remaining parameters. COBRA incorporates models for both isolated and accelerated systems, as well as both Keplerian and relativistic binaries, and exploits pulse phase information to combine search epochs coherently, over time, frequency or across multiple telescopes. We demonstrate the efficacy of our approach in a series of simulations that challenge typical search techniques, including highly aliased signals, and relativistic binary systems. In the most extreme case, we simulate an 8 h observation containing 24 orbits of a pulsar in a binary with a 30 M⊙ companion. Even in this scenario we show that we can build up from an initial low-significance candidate, to fully recovering the signal. We also apply the method to survey data of three pulsars from the globular cluster 47Tuc: PSRs J0024-7204D, J0023-7203J and J0024-7204R. This final pulsar is in a 1.6 h binary, the shortest of any pulsar in 47Tuc, and additionally shows significant scintillation. By allowing the amplitude of the source to vary as a function of time, however, we show that we are able to obtain optimal combinations of such noisy data. We also demonstrate the ability of COBRA to perform high-precision pulsar timing directly on the single pulse survey data, and obtain a 95 per cent upper limit on the eccentricity of PSR J0024-7204R of εb < 0.0007.

Validity and reliability of the Cognitive Complaints in Bipolar Disorder Rating Assessment (COBRA) in Japanese patients with bipolar disorder.

Science.gov (United States)

Toyoshima, Kuniyoshi; Fujii, Yutaka; Mitsui, Nobuyuki; Kako, Yuki; Asakura, Satoshi; Martinez-Aran, Anabel; Vieta, Eduard; Kusumi, Ichiro

2017-08-01

In Japan, there are currently no reliable rating scales for the evaluation of subjective cognitive impairment in patients with bipolar disorder. We studied the relationship between the Japanese version of the Cognitive Complaints in Bipolar Disorder Rating Assessment (COBRA) and objective cognitive assessments in patients with bipolar disorder. We further assessed the reliability and validity of the COBRA. Forty-one patients, aged 16-64, in a remission period of bipolar disorder were recruited from Hokkaido University Hospital in Sapporo, Japan. The COBRA (Japanese version) and Frankfurt Complaint Questionnaire (FCQ), the gold standard in subjective cognitive assessment, were administered. A battery of neuropsychological tests was employed to measure objective cognitive impairment. Correlations among the COBRA, FCQ, and neuropsychological tests were determined using Spearman's correlation coefficient. The Japanese version of the COBRA had high internal consistency, good retest reliability, and concurrent validity-as indicated by a strong correlation with the FCQ. A significant correlation was also observed between the COBRA and objective cognitive measurements of processing speed. These findings are the first to demonstrate that the Japanese version of the COBRA may be clinically useful as a subjective cognitive impairment rating scale in Japanese patients with bipolar disorder. Copyright © 2017 Elsevier Ireland Ltd. All rights reserved.

Designing the User Interface COBRET under Windows to Carry out Pre- and Post-Processing for the Programs COBRA-RERTR and PARET

International Nuclear Information System (INIS)

Ghazi, N.; Monther, A.; Hainoun, A.

2004-01-01

In the frame work of testing, evaluation and application of computer codes in the design and safety analysis of research reactors, the dynamic code PARET and the thermal hydraulic code COBRA-RERTR have been adopted. In order to run the codes under windows and to support the user by pre- and post processing, the user interface program COBRET has been developed in the programming language Visual Basic 6 and the data used by it are organized and stored in a relational database in MS Access, an integral part of the software package, MS Office. The interface works in the environment of the Windows operating system and utilizes its graphics as well as other possibilities. It consists of Pre and Post processor. The pre processor deals with the interactive preparation of the input files for PARET and COBRA codes. It supports the user with an automatic check in routine for detecting logical input errors in addition to many direct helps during the multi mode input process. This process includes an automatic branching according to the selected control parameters that depends on the simulation modes of the considered physical problem. The post processor supports the user with graphical tool to present the time and axial distribution of the system variables that consist of many neutronics and thermal hydraulic parameters of the reactor system like neutron flux, reactivity, temperatures, flow rate, pressure and void distribution. (authors)

Designing the user interface COBRET under windows to carry out pre- and post-processing for the programs COBRA-RERTR and PARET

International Nuclear Information System (INIS)

Hainoun, A.; Monther, A.; Ghazi, N.

2004-01-01

In this framework of testing, evaluation and application of computer codes in the design studies and safety analysis of research reactors, the dynamic code PARET and the thermal hydraulic code COBRA-RERTR have been adopted. In order to run the codes under windows and to support the user by pre- and post processing. The user interface program COBRET has been developed in the programming language visual basic 6 and the data used by it are organized and stored in a relational database in MS Access, and integral port of the software package, MS Office. The interface works in the environment of the Windows operating system and utilizes its graphics as well as other possibilities. It consists of Pre and Post processor. The pre processor deals with the interactive preparation of the input files for PARET and COBRA codes. it supports the user with an automatic check in routine for detecting logical input errors in addition to many direct helps during the multi mode input process. This process includes an automatic branching according to the selected control parameters that depends on the simulation modes of the considered physical problem. The post processor supports the user with graphical tool to present the time and axial distribution of the system variables that consist of many neutronics and thermal hydraulic parameters o the reactor system like neutron flux, reactivity, temperatures, flow rate, pressure and void distribution (author)

COBRA-Bee Carpal-Wrist Gimbal for Astrobee, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — TUI proposes to develop a carpal-wrist gimbal payload for the Astrobee free-flier, called 'COBRA-Bee' to satisfy Astrobee mission needs for a lightweight, integrated...

The popsicle-stick cobra wave

OpenAIRE

Boucher , Jean-Philippe; Clanet , Christophe; Quéré , David; Chevy , Frédéric

2017-01-01

The cobra wave is a popular physical phenomenon arising from the explosion of a metastable grillage made of popsicle sticks. The sticks are expelled from the mesh by releasing the elastic energy stored during the weaving of the structure. Here we analyse both experimentally and theoretically the propagation of the wave-front depending on the properties of the sticks and the pattern of the mesh. We show that its velocity and its shape are directly related to the recoil imparted to the structur...

A study of ribonuclease activity in venom of vietnam cobra

Directory of Open Access Journals (Sweden)

Thiet Van Nguyen

2017-09-01

Full Text Available Abstract Background Ribonuclease (RNase is one of the few toxic proteins that are present constantly in snake venoms of all types. However, to date this RNase is still poorly studied in comparison not only with other toxic proteins of snake venom, but also with the enzymes of RNase group. The objective of this paper was to investigate some properties of RNase from venom of Vietnam cobra Naja atra. Methods Kinetic methods and gel filtration chromatography were used to investigate RNase from venom of Vietnam cobra. Results RNase from venom of Vietnam cobra Naja atra has some characteristic properties. This RNase is a thermostable enzyme and has high conformational stability. This is the only acidic enzyme of the RNase A superfamily exhibiting a high catalytic activity in the pH range of 1–4, with pHopt = 2.58 ± 0.35. Its activity is considerably reduced with increasing ionic strength of reaction mixture. Venom proteins are separated by gel filtration into four peaks with ribonucleolytic activity, which is abnormally distributed among the isoforms: only a small part of the RNase activity is present in fractions of proteins with molecular weights of 12–15 kDa and more than 30 kDa, but most of the enzyme activity is detected in fractions of polypeptides, having molecular weights of less than 9 kDa, that is unexpected. Conclusions RNase from the venom of Vietnam cobra is a unique member of RNase A superfamily according to its acidic optimum pH (pHopt = 2.58 ± 0.35 and extremely low molecular weights of its major isoforms (approximately 8.95 kDa for RNase III and 5.93 kDa for RNase IV.

COBRA-SFS predictions of single assembly spent fuel heat transfer data

International Nuclear Information System (INIS)

Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.; Rector, D.R.

1986-04-01

The study reported here is one of several efforts to evaluate and qualify the COBRA-SFS computer code for use in spent fuel storage system thermal analysis. The ability of COBRA-SFS to predict the thermal response of two single assembly spent fuel heat transfer tests was investigated through comparisons of predictions with experimental test data. From these comparisons, conclusions regarding the computational treatment of the physical phenomena occurring within a storage system can be made. This objective was successfully accomplished as reasonable agreement between predictions and data were obtained for the 21 individual test cases of the two experiments

Purification and antibacterial activities of an L-amino acid oxidase from king cobra (Ophiophagus hannah venom

Directory of Open Access Journals (Sweden)

CS Phua

2012-01-01

Full Text Available Some constituents of snake venom have been found to display a variety of biological activities. The antibacterial property of snake venom, in particular, has gathered increasing scientific interest due to antibiotic resistance. In the present study, king cobra venom was screened against three strains of Staphylococcus aureus [including methicillin-resistant Staphylococcus aureus (MRSA], three other species of gram-positive bacteria and six gram-negative bacteria. King cobra venom was active against all the 12 bacteria tested, and was most effective against Staphylococcus spp. (S. aureus and S. epidermidis. Subsequently, an antibacterial protein from king cobra venom was purified by gel filtration, anion exchange and heparin chromatography. Mass spectrometry analysis confirmed that the protein was king cobra L-amino acid oxidase (Oh-LAAO. SDS-PAGE showed that the protein has an estimated molecular weight of 68 kDa and 70 kDa under reducing and non-reducing conditions, respectively. The minimum inhibitory concentrations (MIC of Oh-LAAO for all the 12 bacteria were obtained using radial diffusion assay method. Oh-LAAO had the lowest MIC value of 7.5 µg/mL against S. aureus ATCC 25923 and ATCC 29213, MRSA ATCC 43300, and S. epidermidis ATCC 12228. Therefore, the LAAO enzyme from king cobra venom may be useful as an antimicrobial agent.

[Spitting cobras: description of 2 cases in Djibouti].

Science.gov (United States)

Rouvin, B; Kone, M; N'diaye, M; Seck, M; Diatta, B

2010-02-01

The purpose of this report is to describe two cases involving ophthalmic exposure to venom from spitting cobras. Based on these cases, readers are reminded that eye injury can be prevented by low-cost treatment consisting of prompt, prolonged saline irrigation. This treatment also reduces pain.

Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

International Nuclear Information System (INIS)

Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

2005-01-01

Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

Parallelized preconditioned BiCGStab solution of sparse linear system equations in F-COBRA-TF

International Nuclear Information System (INIS)

Geemert, Rene van; Glück, Markus; Riedmann, Michael; Gabriel, Harry

2011-01-01

Recently, the in-house development of a preconditioned and parallelized BiCGStab solver has been pursued successfully in AREVA’s advanced sub-channel code F-COBRA-TF. This solver can be run either in a sequential computation mode on a single CPU, or in a parallel computation mode on multiple parallel CPUs. The developed procedure enables the computation of several thousands of successive sparse linear system solutions in F-COBRA-TF with acceptable wall clock run times. The current paper provides general information about F-COBRA-TF in terms of modeling capabilities and application areas, and points out where the relevance arises for the efficient iterative solution of sparse linear systems. Furthermore, the preconditioning and parallelization strategies in the developed BiCGStab iterative solution approach are discussed. The paper is concluded with a number of verification examples. (author)

Cobra-TF simulation of BWR bundle dry out experiments

Energy Technology Data Exchange (ETDEWEB)

Frepoli, C.; Ireland, A.; Hochreiter, L.; Ivanov, K. [Penn State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, PA (United States); Velten, R. [Siemens Nuclear Power GmbH, Erlangen (Germany)

2001-07-01

The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. This study presents the results obtained with COBRA-TF for the simulation of the Siemens 9-9Q BWR Bundle Dryout experiments. The model includes 20 channels and 34 axial nodes in the heated section. The predicted critical power and dryout location is compared with the measured values. An assessment of the code entrainment and de-entrainment models is presented. (authors)

Venom-gland transcriptome and venom proteome of the Malaysian king cobra (Ophiophagus hannah).

Science.gov (United States)

Tan, Choo Hock; Tan, Kae Yi; Fung, Shin Yee; Tan, Nget Hong

2015-09-10

The king cobra (Ophiophagus hannah) is widely distributed throughout many parts of Asia. This study aims to investigate the complexity of Malaysian Ophiophagus hannah (MOh) venom for a better understanding of king cobra venom variation and its envenoming pathophysiology. The venom gland transcriptome was investigated using the Illumina HiSeq™ platform, while the venom proteome was profiled by 1D-SDS-PAGE-nano-ESI-LCMS/MS. Transcriptomic results reveal high redundancy of toxin transcripts (3357.36 FPKM/transcript) despite small cluster numbers, implying gene duplication and diversification within restricted protein families. Among the 23 toxin families identified, three-finger toxins (3FTxs) and snake-venom metalloproteases (SVMPs) have the most diverse isoforms. These 2 toxin families are also the most abundantly transcribed, followed in descending order by phospholipases A2 (PLA2s), cysteine-rich secretory proteins (CRISPs), Kunitz-type inhibitors (KUNs), and L-amino acid oxidases (LAAOs). Seventeen toxin families exhibited low mRNA expression, including hyaluronidase, DPP-IV and 5'-nucleotidase that were not previously reported in the venom-gland transcriptome of a Balinese O. hannah. On the other hand, the MOh proteome includes 3FTxs, the most abundantly expressed proteins in the venom (43 % toxin sbundance). Within this toxin family, there are 6 long-chain, 5 short-chain and 2 non-conventional 3FTx. Neurotoxins comprise the major 3FTxs in the MOh venom, consistent with rapid neuromuscular paralysis reported in systemic envenoming. The presence of toxic enzymes such as LAAOs, SVMPs and PLA2 would explain tissue inflammation and necrotising destruction in local envenoming. Dissimilarities in the subtypes and sequences between the neurotoxins of MOh and Naja kaouthia (monocled cobra) are in agreement with the poor cross-neutralization activity of N. kaouthia antivenom used against MOh venom. Besides, the presence of cobra venom factor, nerve growth factors

Background simulation for the COBRA-experiment

Energy Technology Data Exchange (ETDEWEB)

Quante, Thomas [TU Dortmund, Institut fuer Physik (Germany); Collaboration: COBRA-Collaboration

2015-07-01

COBRA is a next-generation experiment searching for neutrinoless double beta (0νββ) decay using CdZnTe semiconductor detectors. The main focus is on {sup 116}Cd, with a Q-value of 2813.5 keV well above the highest dominant naturally occurring gamma lines. By measuring the half-life of the 0νββ decay, it is possible to clarify the nature of the neutrino as either Dirac or Majorana particle and furthermore to determine the effective Majorana mass. COBRA is currently in the demonstrator phase to study possible background contributions and gain information about the longterm stability of the used detectors. For this purpose a demonstrator array made up of 64 Cadmium-Zinc-Telluride (CdZnTe) semiconductor detectors in coplanar grid configuration was designed and realised at the Gran Sasso Underground laboratory (LNGS) in Italy. Simulations of the whole demonstrator setup are ongoing to reproduce the measured spectra for each detector. This is done in two steps. The first uses the Geant4 based framework VENOM for tracking and energy deposition inside each detector. Detector effects like the energy resolution and electron trapping have to be applied in the second step. The used detector geometry has to be verified against calibration measurements. This talk gives an overview of the current simulation status.

Transient Analysis of Generation IV quick reactors; Analisis de Transitorios en Reactores Rapidos de Generacion IV

Energy Technology Data Exchange (ETDEWEB)

Vazquez, M.; Martin-Fuertes, F.

2013-07-01

As a complement to the attached code 3D neutron-CIEMAT thermohydraulic added a module to simulate transient. Temporary kinetics is resolved by factoring flow in a spatial part and another storm. MCNP provides the reactivity and updated spatial function and COBRA-IV calculates the temperature distribution. Temporary dependence of amplitude is calculated using time delayed neutron Kinetic equations. As an example of application, examines a transient loss of flow in MYRRHA, a lead-cooled experimental reactor.

Qualification of the coupled RELAP5/PANTHER/COBRA code package for licensing applications

International Nuclear Information System (INIS)

Schneidesch, C.R.; Zhang Jinzhao

2004-01-01

A coupled thermal hydraulics-neutronics code package has been developed at Tractebel Engineering (TE), in which the best-estimate thermal-hydraulic system code, RELAP5/mod2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via the dynamic data exchange interface, TALINK. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by the sub-channel thermal-hydraulic analysis code COBRA-3C. The package provides the capability to accurately simulate the key physical phenomena in nuclear power plant accidents with strong asymmetric behaviours and system-core interactions. This paper presents the TE coupled code package and focuses on the methodology followed for qualifying it for licensing applications. The qualification of the coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric PWR accidents with strong core-system interactions

Analyses of venom spitting in African cobras (Elapidae: Serpentes ...

African Journals Online (AJOL)

... all four species. The low levels of variation in venom volume, coupled with the variation in venom dispersal pattern, suggests a complexity to the regulation of venom flow in spitting cobras beyond simply neuromuscular control of the extrinsic venom gland. Keywords: defensive behaviour, snake, teeth, Naja, Hemachatus ...

Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

International Nuclear Information System (INIS)

Umbehaun, Pedro Ernesto

2000-01-01

This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

Introducing COBRAs: exploring motivations for brand-related social media use

NARCIS (Netherlands)

Muntinga, D.G.; Moorman, M.; Smit, E.G.

2011-01-01

The article examines the use of social media by Internet users related to advertising and marketing, called "consumers' online brand-related activities (COBRA)." Interviews are conducted with such Internet users through instant messaging as to their motivations for engaging with brands and brand

Quercetin modulates activities of Taiwan cobra phospholipase A 2 ...

Indian Academy of Sciences (India)

Home; Journals; Journal of Biosciences; Volume 37; Issue 2. Quercetin modulates activities of Taiwan cobra phospholipase A2 via its effects on membrane structure and membrane-bound mode of phospholipase A2. Yi-Ling Chiou Shinne-Ren Lin Wan-Ping Hu Long-Sen Chang. Articles Volume 37 Issue 2 June 2012 pp ...

COBRA 9121: Federal liability for patient screening and transfer.

Science.gov (United States)

Frew, S A

1988-01-01

Health care is no longer a simple cottage industry of individual providers. Increases in competition and government regulation have transformed the old structure of health care into a fend-for-yourself marketplace dominated by multi-institutional corporations. In order to accomplish this change, health care providers have had to alter their locus of attention from the patient to the bottom line. As a result, it is not surprising to find corporate business practices interspersed among the traditional health care practices. On March 1, 1987, the federal government began an assault on a casualty of this new market oriental philosophy, patient transfers or "dumping". COBRA 9121 is an "anti-dumping" law designed to prevent hospitals from continuing this practice. The vehicle for ensuring that the statute's broad provisions are followed is a set of "sudden death" probations. For example, under COBRA, hospitals found guilty of knowing or negligent violations may be suspended or terminated from receiving all Medicare reimbursement. One way to avoid these "sudden death" probations is to understand the implications of this law.

COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual

International Nuclear Information System (INIS)

Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.

1986-11-01

COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code

Comparison of the air-Q intubating laryngeal airway and the cobra perilaryngeal airway as conduits for fiber optic-guided intubation in pediatric patients.

Science.gov (United States)

Girgis, Karim K; Youssef, Maha M I; ElZayyat, Nashwa S

2014-10-01

One of the methods proposed in cases of difficult airway management in children is using a supraglottic airway device as a conduit for tracheal intubation. The aim of this study was to compare the efficacy of the Air-Q Intubating Laryngeal Airway (Air-Q) and the Cobra Perilaryngeal Airway (CobraPLA) to function as a conduit for fiber optic-guided tracheal intubation in pediatric patients. A total of 60 children with ages ranging from 1 to 6 years, undergoing elective surgery, were randomized to have their airway managed with either an Air-Q or CobraPLA. Outcomes recorded were the success rate, time and number of attempts required for fiber optic-guided intubation and the time required for device removal after intubation. We also recorded airway leak pressure (ALP), fiber optic grade of glottic view and occurrence of complications. Both devices were successfully inserted in all patients. The intubation success rate was comparable with the Air-Q and the CobraPLA (96.7% vs. 90%), as was the first attempt success rate (90% vs. 80%). The intubation time was significantly longer with the CobraPLA (29.5 ± 10.9 s vs. 23.2 ± 9.8 s; P fiber optic grade of glottic view was comparable with the two devices. The CobraPLA was associated with a significantly higher incidence of blood staining of the device on removal and post-operative sore throat. Both the Air-Q and CobraPLA can be used effectively as a conduit for fiber optic-guided tracheal intubation in children. However, the Air-Q proved to be superior due to a shorter intubation time and less airway morbidity compared with the CobraPLA.

Dynamic power behavior of a PWR type nuclear reactor

International Nuclear Information System (INIS)

Moreira, F.J.

1984-01-01

A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.) [pt

Development of Brigade Staff Tasks for the COBRAS II Brigade Staff Exercise

National Research Council Canada - National Science Library

Deter, Daniel

1998-01-01

... and development of simulation-based training for the conventional mounted brigade staff. The work was performed under a project called Combined Arms Operations at Brigade Level, Realistically Achieved Through Simulation (COBRAS).

Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

International Nuclear Information System (INIS)

Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

2005-01-01

Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

Energy Technology Data Exchange (ETDEWEB)

Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

2005-07-01

Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)

International Nuclear Information System (INIS)

Burns, C. J.; Aumiler, D.L.

2006-01-01

The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF

Evaluation of the use of metal alloy fuels in pressurized water reactors

International Nuclear Information System (INIS)

1990-01-01

The project concentrated on model development. Reactor physics modeling involved establishing accurate models with PC versions of COMBINE and VENTURE. Fuel performance analysis will start with METAL- LIFE. In order to justify the change of fuel to metal alloy, large benefits will have to be found; the cost benefit reported is not sufficient. The fuel pin will be annular and contact the clad; the clad thickness will force the fuel to grow toward the central hole. This report reports: design improvements, neutronic model development, COBRA modifications, reactor kinetics model development, RELAP code, and fuel performance

Evaluation and enhancement of COBRA-TF efficiency for LWR calculations

International Nuclear Information System (INIS)

Cuervo, Diana; Avramova, Maria; Ivanov, Kostadin; Miro, Rafael

2006-01-01

Detailed representations of the reactor core generate computational meshes with a high number of cells where the fluid dynamics equations must be solved. An exhaustive analysis of the CPU times needed by the thermal-hydraulic subchannel code COBRA-TF for different stages in the solution process has revealed that the solution of the linear system of pressure equations is the most time consuming process. To improve code efficiency two optimized matrix solvers, Super LU library and Krylov non-stationary iterative methods have been implemented in the code and their performance has been tested using a suite of five test cases. The results of performed comparative analyses have demonstrated that for large cases, the implementation of the Bi-Conjugate Gradient Stabilized (Bi-CGSTAB) Krylov method combined with the incomplete LU factorization with dual truncation strategy (ILUT) pre-conditioner reduced the time used by the code for the solution of the pressure matrix by a factor of 20. Both new solvers converge smoothly regardless of the nature of simulated cases and the mesh structures and improve the stability and accuracy of results compared to the classic Gauss-Seidel iterative method. The obtained results indicate that the direct inversion method is the best option for small cases

CoBra - a global tool for braking system development; CoBra - ein Tool fuer den globalen Einsatz in der Bremssystementwicklung

Energy Technology Data Exchange (ETDEWEB)

Sailer, U. [Robert Bosch GmbH, Stuttgart (Germany)

1999-07-01

When Robert Bosch GmbH took over the braking system activities of Allied Signal in 1996, they became able to develop complete braking systems for passenger cars. Braking systes for the different markets are now produced in three sites in Germany, France and the USA. Braking system development with its interfaces to component development and to car producers is a new development challenge, and the department K1 (ABS and braking systems) is cooperating with internal and external partners in developing a globally standardized program for design, simulation and analysis of passenger car braking systems. This contribution presents parts of the development project CoBra (Computation of Braking Systems). [German] Mit dem Kauf der Bremsenaktivitaeten der Firma Allied Signal im Jahre 1996 ist die Robert Bosch GmbH in der Lage, komplette Pkw-Bremssysteme zu entwickeln. Nunmehr werden an drei Entwicklungsstandorten in Deutschland, Frankreich und den USA Bremssysteme fuer die verschiedenen Maerkte entwickelt. Die Bremssystementwicklung, insbesondere die damit verbundenen Schnittstellen zu der Komponentenentwicklung und zum Automobilhersteller, stellt technisch und vom Entwicklungsprozess aus gesehen eine neue Herausforderung dar. Um ihr zu begegnen, wird im Geschaeftsbereich K1 (ABS and Braking Systems) derzeit in Zusammenarbeit mit internen und externen Partnern ein global einheitliches Programm zur Auslegung, Simulation und Analyse von Pkw-Bremssystemen entwickelt. Dieser Beitrag stellt Teile des Entwicklungsprojekts CoBra (Computation of Braking Systems) vor. (orig.)

Reactor operational transient analysis

International Nuclear Information System (INIS)

Shin, W.K.; Chae, S.K.; Han, K.I.; Yang, K.S.; Chung, H. D.; Kim, H.G.; Moon, H.J.; Ryu, Y.H.

1983-01-01

To build up efficient capability of safety review and inspection for the nuclear power plants, four area of studies have performed as follows: 1) In order to search the most optimized operating method during load follow operating schemes, automatic control and normal control, are compared each other under the CAOC condition. The analysis performed by DDID code has shown that the reactor has to be controlled by the operator manually during load follow operation. 2) Through the sensitivity analysis by COBRA code, the operating parameters, such as coolant pressure, flow rate, inlet temperature, and power distribution are shown to be important to the determination of DNBR. Expecially, inlet temperature of primary coolant system is appeared as the most senstive parameter on DNBR. 3) FRAPCON code is adapted to study the sensitivity of several operational parameters on the mechanical properties of reactor fuel rod. 4) The calculations procedure which is required to be obtained the neutron fluence at the reactor vessel and the spectrum at the surveillance capsule is established. The results of computation are conpared with those of FSAR and SWRI report and proved its applicability to reactor surveillance program. (Author)

Study of gas-puff Z-pinches on COBRA

Energy Technology Data Exchange (ETDEWEB)

Qi, N.; Rosenberg, E. W.; Gourdain, P. A.; Grouchy, P. W. L. de; Kusse, B. R.; Hammer, D. A.; Bell, K. S.; Shelkovenko, T. A.; Potter, W. M.; Atoyan, L.; Cahill, A. D.; Evans, M.; Greenly, J. B.; Hoyt, C. L.; Pikuz, S. A.; Schrafel, P. C. [Laboratory of Plasma Studies, Cornell University, Ithaca, New York 14853 (United States); Kroupp, E.; Fisher, A.; Maron, Y. [Weizmann Institute of Science, Rehovot 76100 (Israel)

2014-11-15

Gas-puff Z-pinch experiments were conducted on the 1 MA, 200 ns pulse duration Cornell Beam Research Accelerator (COBRA) pulsed power generator in order to achieve an understanding of the dynamics and instability development in the imploding and stagnating plasma. The triple-nozzle gas-puff valve, pre-ionizer, and load hardware are described. Specific diagnostics for the gas-puff experiments, including a Planar Laser Induced Fluorescence system for measuring the radial neutral density profiles along with a Laser Shearing Interferometer and Laser Wavefront Analyzer for electron density measurements, are also described. The results of a series of experiments using two annular argon (Ar) and/or neon (Ne) gas shells (puff-on-puff) with or without an on- (or near-) axis wire are presented. For all of these experiments, plenum pressures were adjusted to hold the radial mass density profile as similar as possible. Initial implosion stability studies were performed using various combinations of the heavier (Ar) and lighter (Ne) gasses. Implosions with Ne in the outer shell and Ar in the inner were more stable than the opposite arrangement. Current waveforms can be adjusted on COBRA and it was found that the particular shape of the 200 ns current pulse affected on the duration and diameter of the stagnated pinched column and the x-ray yield.

COBRA: A Computational Brewing Application for Predicting the Molecular Composition of Organic Aerosols

Energy Technology Data Exchange (ETDEWEB)

Fooshee, David R.; Nguyen, Tran B.; Nizkorodov, Sergey A.; Laskin, Julia; Laskin, Alexander; Baldi, Pierre

2012-05-08

Atmospheric organic aerosols (OA) represent a significant fraction of airborne particulate matter and can impact climate, visibility, and human health. These mixtures are difficult to characterize experimentally due to the enormous complexity and dynamic nature of their chemical composition. We introduce a novel Computational Brewing Application (COBRA) and apply it to modeling oligomerization chemistry stemming from condensation and addition reactions of monomers pertinent to secondary organic aerosol (SOA) formed by photooxidation of isoprene. COBRA uses two lists as input: a list of chemical structures comprising the molecular starting pool, and a list of rules defining potential reactions between molecules. Reactions are performed iteratively, with products of all previous iterations serving as reactants for the next one. The simulation generated thousands of molecular structures in the mass range of 120-500 Da, and correctly predicted ~70% of the individual SOA constituents observed by high-resolution mass spectrometry (HR-MS). Selected predicted structures were confirmed with tandem mass spectrometry. Esterification and hemiacetal formation reactions were shown to play the most significant role in oligomer formation, whereas aldol condensation was shown to be insignificant. COBRA is not limited to atmospheric aerosol chemistry, but is broadly applicable to the prediction of reaction products in other complex mixtures for which reasonable reaction mechanisms and seed molecules can be supplied by experimental or theoretical methods.

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

International Nuclear Information System (INIS)

Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

1986-01-01

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

Development in High-Density Cobra Fiber Positioners for the Subaru Telescope's Prime Focus Spectrometer

Science.gov (United States)

Fisher, Charles D.; Braun, David F.; Kaluzny, Joel V.; Seiffert, Mic D.; Dekany, Richard G.; Ellis, Richard S.; Smith, Roger S.

2012-01-01

The Prime Focus Spectrograph (PFS) is a fiber fed multi-object spectrometer for the Subaru Telescope that will conduct a variety of targeted surveys for studies of dark energy, galaxy evolution, and galactic archaeology. The key to the instrument is a high density array of fiber positioners placed at the prime focus of the Subaru Telescope. The system, nicknamed "Cobra", will be capable of rapidly reconfiguring the array of 2394 optical fibers to the image positions of astronomical targets in the focal plane with high accuracy. The system uses 2394 individual "SCARA robot" mechanisms that are 7.7mm in diameter and use 2 piezo-electric rotary motors to individually position each of the optical fibers within its patrol region. Testing demonstrates that the Cobra positioner can be moved to within 5 micrometers of an astronomical target in 6 move iterations with a success rate of 95%. The Cobra system is a key aspect of PFS that will enable its unprecedented combination of high-multiplex factor and observing efficiency on the Subaru telescope. The requirements, design, and prototyping efforts for the fiber positioner system for the PFS are described here as are the plans for modular construction, assembly, integration, functional testing, and performance validation.

Blindness from spitting cobra venom: Case report | Atipo-Tsiba | East ...

African Journals Online (AJOL)

Spitting cobra is the name given to some snakes of the family of Elapidae, belonging to the genus Naja or Hemachatus that have the ability to spitt heir venom (up to 3m) to blind their predators. Naja mossambica is the most answered species in Africa.The precise statistics of attacks due to this snake are available, let alone ...

High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

International Nuclear Information System (INIS)

Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

2013-01-01

The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

Introducing COBRAs: a holistic exploration of motivations for brand-related social media use

NARCIS (Netherlands)

Muntinga, D.G.; Moorman, M.; Smit, E.G.

2009-01-01

Propelled by highly interactive technologies, internet users are increasingly becoming active consumers, contributors and producers of content on brands. Consumer’s Online Brand-Related Activities ("COBRAs") have significant consequences for firms and brands. To effectively respond to and steer

Utilization of a statistical procedure for DNBR calculation and in the survey of reactor protection limits

International Nuclear Information System (INIS)

Pontedeiro, A.C.; Camargo, C.T.M.; Galetti, M.R. da Silva.

1987-01-01

A new procedure is applied to Angra 1 NPP, which is related to DNBR calculations, considering the design parameters statistically: Improved Thermal Design Procedure (ITDP). The ITDP application leads to the determination of uncertainties in the input parameters, the sensitivity factors on DNBR. The DNBR limit and new reactor protection limits. This was done to Angra 1 with the subchannel code COBRA-IIIP. The analysis of limiting accident in terms of DNB confirmed a gain in DNBR margin, and greater operation flexibility of the plant, decreasing unnecessary trips of the reactor. (author) [pt

The action of cobra venom phospholipase A2 isoenzymes towards intact human erythrocytes

NARCIS (Netherlands)

Roelofsen, B.; Sibenius Trip, M.; Verheij, H.M.; Zevenbergen, J.L.

1980-01-01

1. 1. Cobra venom phospholipase A2 from three different sources has been fractionated into different isoenzymes by DEAE ion-exchange chromatography. 2. 2. Treatment of intact human erythrocytes with the various isoenzymes revealed significant differences in the degree of phosphatidylcholine

Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

International Nuclear Information System (INIS)

Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

2002-01-01

A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

PENSAMIENTOS COMPARTIDOS. ALDO VAN EYCK, EL GRUPO COBRA Y EL ARTE / Shared thoughts. Aldo van Eyck, the COBRA group, and art

Directory of Open Access Journals (Sweden)

Esther Mayoral Campa

2014-11-01

Full Text Available RESUMEN El periodo inmediatamente posterior a la II Guerra Mundial es uno de los episodios más interesantes desde el punto de vista cultural del siglo XX, un momento vivido por muchos de los intelectuales europeos coetáneos a esta época como un punto de inflexión, una oportunidad para repensar el mundo, para comenzar de nuevo tras el cataclismo bélico. En ese contexto comienza su andadura como arquitecto Aldo van Eyck, así como su colaboración con el breve, pero intenso, movimiento Cobra, grupo esencial para comprender el panorama cultural europeo de posguerra y una de las últimas vanguardias del siglo XX. Este artículo explora la vinculación del arquitecto holandés Aldo van Eyck con el mundo del arte. Una relación poliédrica, parte esencial de su discurso, que engloba su formación cultural, sus relaciones de amistad, su pensamiento crítico y su obra. En esa correlación entre la arquitectura y las artes será determinante la vinculación del arquitecto con Cobra, con el que compartirá una mirada común sobre la realidad, una relación compleja con líneas de investigación comunes, escritos, exposiciones y trabajos compartidos. A todo ello se suma la aportación fundamental que supone un trasvase de valores constantes entre la arquitectura y el mundo del arte, que caracterizó la relación entre el arquitecto y los miembros del grupo.

Preliminary study of the thermo-hydraulic behaviour of the binary breeder reactor

International Nuclear Information System (INIS)

Silveira Luz, M. da; Ferreira, W.J.

1984-06-01

Continuing the development of the Binary Breeder Reactor, its physical configuration and the advantages of differents types of spacers are analysed. In order to simulate the thermo-hydraulic behaviour and obtain data for a preliminary evaluation of the core geometry, the COBRA III C code was used to study the effects of the lenght and diameter of the fuel element, the coolant inlet temperature, the system pressure, helicoidal pitch and the pitch to diameter ratio. (Author) [pt

Cognitive effects of electro-acupuncture and pregabalin in a trigeminal neuralgia rat model induced by cobra venom

Directory of Open Access Journals (Sweden)

Chen RW

2017-08-01

Full Text Available Ruo-Wen Chen,1,2 Hui Liu,2 Jian-Xiong An,1,2 Xiao-Yan Qian,2 Yi-De Jiang,2 Doris K Cope,3 John P Williams,3 Rui Zhang,1 Li-Na Sun1 1Department of Anesthesiology, Weifang Medical University, Weifang City, Shandong, 2Department of Anesthesiology, Pain Medicine and Critical Care Medicine, Aviation General Hospital of China Medical University and Beijing Institute of Translational Medicine, Chinese Academy of Sciences, Beijing, China; 3Department of Anesthesiology, University of Pittsburgh School of Medicine, Pittsburgh, PA, USA Objective: The objective of this study was to investigate the effects of electro-acupuncture (EA and pregabalin on cognition impairment induced by chronic trigeminal neuralgia (TN in rats. Design: Controlled animal study. Setting: Department of Anesthesiology, Pain Medicine and Critical Care Medicine, Aviation General Hospital of China Medical University. Subjects: Forty adult male Sprague Dawley rats. Methods: Rats were randomly divided into four groups. The TN model was induced by administration of cobra venom to the left infraorbital nerve. On postoperative day 14, either EA or pregabalin was administered, free behavioral activities were observed. Spatial learning and memory abilities were determined in the Morris water maze. The ultrastructural alterations of the Gasserian ganglion, medulla oblongata and hippocampus were examined by electron microscopy. The changes on long-term potentiation were investigated. Results: After treatment, the exploratory behavior increased and the grooming behavior decreased (P<0.05 for the EA group and pregabalin group compared with the cobra venom group; moreover, demyelination of neurons in Gasserian ganglion and medulla oblongata was reversed. The number of platform site crossings, the average percentages of time in the target quadrant and the field excitatory postsynaptic potential slopes increased (P<0.05 in the EA group compared to the cobra venom group. However, the pregabalin group

Antiproliferative activity of king cobra (Ophiophagus hannah) venom L-amino acid oxidase.

Science.gov (United States)

Li Lee, Mui; Chung, Ivy; Yee Fung, Shin; Kanthimathi, M S; Hong Tan, Nget

2014-04-01

King cobra (Ophiophagus hannah) venom L-amino acid oxidase (LAAO), a heat-stable enzyme, is an extremely potent antiproliferative agent against cancer cells when compared with LAAO isolated from other snake venoms. King cobra venom LAAO was shown to exhibit very strong antiproliferative activities against MCF-7 (human breast adenocarcinoma) and A549 (human lung adenocarcinoma) cells, with an IC50 value of 0.04±0.00 and 0.05±0.00 μg/mL, respectively, after 72-hr treatment. In comparison, its cytotoxicity was about 3-4 times lower when tested against human non-tumourigenic breast (184B5) and lung (NL 20) cells, suggesting selective antitumour activity. Furthermore, its potency in MCF-7 and A549 cell lines was greater than the effects of doxorubicin, a clinically established cancer chemotherapeutic agent, which showed an IC50 value of 0.18±0.03 and 0.63±0.21 μg/mL, respectively, against the two cell lines. The selective cytotoxic action of the LAAO was confirmed by phycoerythrin (PE) annexin V/7-amino-actinomycin (AAD) apoptotic assay, in which a significant increase in apoptotic cells was observed in LAAO-treated tumour cells than in their non-tumourigenic counterparts. The ability of LAAO to induce apoptosis in tumour cells was further demonstrated using caspase-3/7 and DNA fragmentation assays. We also determined that this enzyme may target oxidative stress in its killing of tumour cells, as its cytotoxicity was significantly reduced in the presence of catalase (a H2O2 scavenger). In view of its heat stability and selective and potent cytotoxic action on cancer cells, king cobra venom LAAO can be potentially developed for treating solid tumours. © 2013 Nordic Association for the Publication of BCPT (former Nordic Pharmacological Society).

Results of a search for neutrinoless double-beta decay using the COBRA demonstrator

Energy Technology Data Exchange (ETDEWEB)

Quante, Thomas; Goessling, Claus; Kroeninger, Kevin [TU Dortmund, Exp. Physik IV, Dortmund (Germany)

2016-07-01

COBRA is an experiment aiming to search for neutrinoless double-beta-decay (0νββ-decay) using CdZnTe semiconductor detectors. The main focus is on {sup 116}Cd, with a Q-value of 2813.5 keV well above the highest dominant naturally occurring gamma lines. By measuring the half-life of the 0νββ-decay, it is possible to clarify the nature of the neutrino as either Dirac or Majorana particle and furthermore to determine its effective Majorana mass. The COBRA collaboration operates a demonstrator to search for these decays at the Laboratori Nazionali del Gran Sasso in Italy. The exposure of 234.7 kg d considered in this analysis was collected between September 2011 and February 2015. The analysis focuses on the decay of the nuclides {sup 114}Cd, {sup 128}Te, {sup 70}Zn, {sup 130}Te and {sup 116}Cd. A Bayesian analysis is performed to estimate the signal strength of 0νββ-decay.

COBRA System Engineering Processes to Achieve SLI Strategic Goals

Science.gov (United States)

Ballard, Richard O.

2003-01-01

The COBRA Prototype Main Engine Development Project was an endeavor conducted as a joint venture between Pratt & Whitney and Aerojet to conduct risk reduction in LOX/LH2 main engine technology for the NASA Space Launch Initiative (SLI). During the seventeen months of the project (April 2001 to September 2002), approximately seventy reviews were conducted, beginning with the Engine Systems Requirements Review (SRR) and ending with the Engine Systems Interim Design Review (IDR). This paper discusses some of the system engineering practices used to support the reviews and the overall engine development effort.

El cuerpo travesti como urdimbre neobarroca y como desecho en la novela Cobra (1972 de Severo Sarduy. // the transvestite body as a neo-baroque litter and weave in Severo Sarduy`s novel Cobra (1972.

Directory of Open Access Journals (Sweden)

Andrés Arteaga.

2008-12-01

Full Text Available In the current work we will analyze the role of the transvestite body as litter and weave in Cuban writer Severo Sarduy's novel Cobra (1972. First, we will present a brief differentiation among Alejo Carpentier's, Lezama Lima’s, and Sarduy's baroque, from some conceptual elements given by Sarduy about the neo-baroque and the body in his essays La Simulación (1982, Escrito sobre un cuerpo (1968 and Barroco (1974. Subsequently, we will show how all along the novel the status of the main character’s body turns into a place of litter, from Jacques Lacan’s conceptualization about the letter (lettre as litter in the seminar Le Sinthome (1975. // En el presente trabajo analizaremos el lugar del cuerpo travesti como desecho y urdimbre en la novela Cobra (1972 del escritor cubano Severo Sarduy. En primer lugar haremos una breve diferenciación entre el barroco de Alejo Carpentier, Lezama Lima y el barroco de Sarduy a partir de algunos elementos conceptuales que da Sarduy sobre el neobarroco y el cuerpo en sus ensayos La Simulación (1982, Escrito sobre un cuerpo (1968 y Barroco (1974. Posteriormente mostraremos, a lo largo de la novela, cómo el estatus del cuerpo del personaje principal de la novela Cobra deviene un lugar de desecho, a partir de la conceptualización de Jacques Lacan sobre la letra (lettre como desecho (litter en el seminario Le Sinthome (1975.

The New Perilaryngeal Airway (CobraPLA™)1 Is as Efficient as the Laryngeal Mask Airway (LMA™)2, But Provides Better Airway Sealing Pressures

Science.gov (United States)

Akça, Ozan; Wadhwa, Anupama; Sengupta, Papiya; Durrani, Jaleel; Hanni, Keith; Wenke, Mary; Yücel, Yüksel; Lenhardt, Rainer; Doufas, Anthony G.; Sessler, Daniel I.

2006-01-01

The Laryngeal Mask Airway (LMA) is a frequently-used efficient airway device, yet it sometimes seals poorly, thus reducing the efficacy of positive-pressure ventilation. The Perilaryngeal Airway (CobraPLA) is a novel airway device with a larger pharyngeal cuff (when inflated). We tested the hypothesis that the CobraPLA was superior to LMA with regard to insertion time and airway sealing pressure and comparable to LMA in airway adequacy and recovery characteristics. After midazolam and fentanyl, 81 ASA I-II outpatients having elective surgery were randomized to receive an LMA or CobraPLA. Anesthesia was induced with propofol (2.5 mg/kg, IV), and the airway inserted. We measured 1) insertion time; 2) adequacy of the airway (no leak at 15-cm-H2O peak pressure or tidal volume of 5 ml/kg); 3) airway sealing pressure; 4) number of repositioning attempts; and 5) sealing quality (no leak at tidal volume of 8 ml/kg). At the end of surgery, gastric insufflation, postoperative sore throat, dysphonia, and dysphagia were evaluated. Data were compared with unpaired t-tests, chi-square tests, or Fisher’s Exact tests; P<0.05 was significant. Patient characteristics, insertion times, airway adequacy, number of repositioning attempts, and recovery were similar in each group. Airway sealing pressure was significantly greater with CobraPLA (23±6 cm H2O) than LMA (18±5 cm H2O, P<0.001). The CobraPLA has insertion characteristics similar to LMA, but better airway sealing capabilities. PMID:15281543

Cross-reactivity and neutralization of Indian King cobra (Ophiophagus hannah) venom by polyvalent and monovalent antivenoms.

Science.gov (United States)

Gowtham, Yashonandana J; Mahadeswaraswamy, Y H; Girish, K S; K, Kemparaju

2014-07-01

The venom of the largest venomous snake, the king cobra (Ophiophagus hannah), is still out of league for the production of therapeutic polyvalent antivenom nor it is characterized immunologically in the Indian subcontinent. In the present study, the king cobra venom is comparatively studied for the cross-reactivity/reactivity and toxicity neutralization by the locally available equine therapeutic polyvalent BSV and VB antivenoms, and monovalent antivenom (OH-IgG) prepared in rabbit. None of the two therapeutic antivenoms procured from two different firms showed any signs of cross-reactivity in terms of antigen-antibody precipitin lines in immunodouble diffusion assay; however, a weak and an insignificant cross-reactivity pattern was observed in ELISA and Western blot studies. Further, both BSV and VB antivenoms failed to neutralize proteolytic, hyaluronidase and phospholipase activities as well as toxic properties such as edema, myotoxicity and lethality of the venom. As expected, OH-IgG showed strong reactivity in immunodouble diffusion, ELISA and in Western blot analysis and also neutralized both enzyme activities as well as the toxic properties of the venom. Thus, the study provides insight into the likely measures that are to be taken in cases of accidental king cobra bites for which the Indian subcontinent is still not prepared for. Copyright © 2014 Elsevier B.V. All rights reserved.

The World Congress on Controversies in Breast Cancer (CoBRA in Melbourne, Australia

Directory of Open Access Journals (Sweden)

Ilana Rabinoff-Sofer

2015-03-01

Full Text Available The World Congress on Controversies in Breast Cancer (CoBRA will take place October 22-24, 2015 in Melbourne, Australia.CoBRA is a concept congress dealing with controversial topics in breast cancer in the format of debates and discussions, allowing ample time for speaker-participant interaction.CLICK HERE for more information

The analysis of thermal-hydraulic performances of nuclear ship reactor

International Nuclear Information System (INIS)

Wakabayashi, Shinshichi; Hamada, Masao

1975-01-01

Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

Monte Carlo-based development of a shield and total background estimation for the COBRA experiment

International Nuclear Information System (INIS)

Heidrich, Nadine

2014-11-01

The COBRA experiment aims for the measurement of the neutrinoless double beta decay and thus for the determination the effective Majorana mass of the neutrino. To be competitive with other next-generation experiments the background rate has to be in the order of 10 -3 counts/kg/keV/yr, which is a challenging criterion. This thesis deals with the development of a shield design and the calculation of the expected total background rate for the large scale COBRA experiment containing 13824 6 cm 3 CdZnTe detectors. For the development of a shield single-layer and multi-layer shields were investigated and a shield design was optimized concerning high-energy muon-induced neutrons. As the best design the combination of 10 cm boron doped polyethylene as outermost layer, 20 cm lead and 10 cm copper as innermost layer were determined. It showed the best performance regarding neutron attenuation as well as (n, γ) self-shielding effects leading to a negligible background rate of less than 2.10 -6 counts/kg/keV/yr. Additionally. the shield with a thickness of 40 cm is compact and costeffective. In the next step the expected total background rate was computed taking into account individual setup parts and various background sources including natural and man-made radioactivity, cosmic ray-induced background and thermal neutrons. Furthermore, a comparison of measured data from the COBRA demonstrator setup with Monte Carlo data was used to calculate reliable contamination levels of the single setup parts. The calculation was performed conservatively to prevent an underestimation. In addition, the contribution to the total background rate regarding the individual detector parts and background sources was investigated. The main portion arise from the Delrin support structure, the Glyptal lacquer followed by the circuit board of the high voltage supply. Most background events originate from particles with a quantity of 99 % in total. Regarding surface events a contribution of 26

Monte Carlo-based development of a shield and total background estimation for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Heidrich, Nadine

2014-11-15

The COBRA experiment aims for the measurement of the neutrinoless double beta decay and thus for the determination the effective Majorana mass of the neutrino. To be competitive with other next-generation experiments the background rate has to be in the order of 10{sup -3} counts/kg/keV/yr, which is a challenging criterion. This thesis deals with the development of a shield design and the calculation of the expected total background rate for the large scale COBRA experiment containing 13824 6 cm{sup 3} CdZnTe detectors. For the development of a shield single-layer and multi-layer shields were investigated and a shield design was optimized concerning high-energy muon-induced neutrons. As the best design the combination of 10 cm boron doped polyethylene as outermost layer, 20 cm lead and 10 cm copper as innermost layer were determined. It showed the best performance regarding neutron attenuation as well as (n, γ) self-shielding effects leading to a negligible background rate of less than 2.10{sup -6} counts/kg/keV/yr. Additionally. the shield with a thickness of 40 cm is compact and costeffective. In the next step the expected total background rate was computed taking into account individual setup parts and various background sources including natural and man-made radioactivity, cosmic ray-induced background and thermal neutrons. Furthermore, a comparison of measured data from the COBRA demonstrator setup with Monte Carlo data was used to calculate reliable contamination levels of the single setup parts. The calculation was performed conservatively to prevent an underestimation. In addition, the contribution to the total background rate regarding the individual detector parts and background sources was investigated. The main portion arise from the Delrin support structure, the Glyptal lacquer followed by the circuit board of the high voltage supply. Most background events originate from particles with a quantity of 99 % in total. Regarding surface events a

COBRAS/SAMBA: The European space mission to map the CBR anisotropy

DEFF Research Database (Denmark)

Bersanelli, M.; Mandolesi, N.; Cesarsky, C.

1996-01-01

COBRAS/SAMBA is an ESA mission designed for extensive, accurate mapping of the anisotropies of the Cosmic Background Radiation, with angular sensitivity from sub-degree scales up to and overlapping with the COBE-DMR resolution. This will allow a fun identification of the primordial density pertur...... perturbations which grew to form the large-scale structures observed in the present universe. Here we present the scientific goals and the key characteristics of the model payload and observation strategy....

Coherent optical systems implemented for business traffic routing and access: the RACE COBRA project (invited)

NARCIS (Netherlands)

Bachus, E.J.; Almeida, T.; Demeester, P.; Depovere, G.F.G.; Ebberg, A.; Rui Ferreira, M.; Khoe, G.D.; Koning, O.; Marsden, R.; Rawsthorne, J.; Wauters, N.

1996-01-01

The RACE COBRA consortium has performed four field trials to verify the suitability of dense WDM combined with heterodyne detection (for which the term CMC, "Coherent Multicarrier" is used) for different network applications. The main advantage in using heterodyne detection is the simple access to

Mucuna pruriens Linn. seed extract pretreatment protects against cardiorespiratory and neuromuscular depressant effects of Naja sputatrix (Javan spitting cobra) venom in rats.

Science.gov (United States)

Fung, Shin Yee; Tan, Nget Hong; Sim, Si Mui; Marinello, Enrico; Guerranti, Roberto; Aguiyi, John Chinyere

2011-04-01

Mucuna pruriens has been used by native Nigerians as a prophylactic for snakebite. The protective effects of M. pruriens seed extract (MPE) were investigated against the pharmacological actions of N. sputatrix (Javan spitting cobra) venom in rats. The results showed that MPE-pretreatment protected against cardiorespiratory and, to a lesser extent, neuromuscular depressant effects of N. sputatrix venom. These may be explained at least in part by the neutralisation of the cobra venom toxins by anti-MPE antibodies elicited by the MPE pretreatment.

COBRA-LIKE2, a member of the glycosylphosphatidylinositol-anchored COBRA-LIKE family, plays a role in cellulose deposition in arabidopsis seed coat mucilage secretory cells.

Science.gov (United States)

Ben-Tov, Daniela; Abraham, Yael; Stav, Shira; Thompson, Kevin; Loraine, Ann; Elbaum, Rivka; de Souza, Amancio; Pauly, Markus; Kieber, Joseph J; Harpaz-Saad, Smadar

2015-03-01

Differentiation of the maternally derived seed coat epidermal cells into mucilage secretory cells is a common adaptation in angiosperms. Recent studies identified cellulose as an important component of seed mucilage in various species. Cellulose is deposited as a set of rays that radiate from the seed upon mucilage extrusion, serving to anchor the pectic component of seed mucilage to the seed surface. Using transcriptome data encompassing the course of seed development, we identified COBRA-LIKE2 (COBL2), a member of the glycosylphosphatidylinositol-anchored COBRA-LIKE gene family in Arabidopsis (Arabidopsis thaliana), as coexpressed with other genes involved in cellulose deposition in mucilage secretory cells. Disruption of the COBL2 gene results in substantial reduction in the rays of cellulose present in seed mucilage, along with an increased solubility of the pectic component of the mucilage. Light birefringence demonstrates a substantial decrease in crystalline cellulose deposition into the cellulosic rays of the cobl2 mutants. Moreover, crystalline cellulose deposition into the radial cell walls and the columella appears substantially compromised, as demonstrated by scanning electron microscopy and in situ quantification of light birefringence. Overall, the cobl2 mutants display about 40% reduction in whole-seed crystalline cellulose content compared with the wild type. These data establish that COBL2 plays a role in the deposition of crystalline cellulose into various secondary cell wall structures during seed coat epidermal cell differentiation. © 2015 American Society of Plant Biologists. All Rights Reserved.

Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

2014-01-01

The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

Effect of Mucuna pruriens Seed Extract Pretreatment on the Responses of Spontaneously Beating Rat Atria and Aortic Ring to Naja sputatrix (Javan Spitting Cobra) Venom

Science.gov (United States)

Fung, Shin Yee; Tan, Nget Hong; Sim, Si Mui; Aguiyi, John C.

2012-01-01

Mucuna pruriens Linn. (velvet bean) has been used by native Nigerians as a prophylactic for snakebite. Rats pretreated with M. pruriens seed extract (MPE) have been shown to protect against the lethal and cardiovascular depressant effects of Naja sputatrix (Javan spitting cobra) venoms, and the protective effect involved immunological neutralization of the venom toxins. To investigate further the mechanism of the protective effect of MPE pretreatment against cobra venom toxicity, the actions of Naja sputatrix venom on spontaneously beating rat atria and aortic rings isolated from both MPE pretreated and untreated rats were studied. Our results showed that the MPE pretreatment conferred protection against cobra venom-induced depression of atrial contractility and atrial rate in the isolated atrial preparations, but it had no effect on the venom-induced contractile response of aortic ring preparation. These observations suggested that the protective effect of MPE pretreatment against cobra venom toxicity involves a direct protective action of MPE on the heart function, in addition to the known immunological neutralization mechanism, and that the protective effect does not involve action on blood vessel contraction. The results also suggest that M. pruriens seed may contain novel cardioprotective agent with potential therapeutic value. PMID:21785646

COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi

1987-01-01

The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)

Antibacterial action of a heat-stable form of L-amino acid oxidase isolated from king cobra (Ophiophagus hannah) venom.

Science.gov (United States)

Lee, Mui Li; Tan, Nget Hong; Fung, Shin Yee; Sekaran, Shamala Devi

2011-03-01

The major l-amino acid oxidase (LAAO, EC 1.4.3.2) of king cobra (Ophiophagus hannah) venom is known to be an unusual form of snake venom LAAO as it possesses unique structural features and unusual thermal stability. The antibacterial effects of king cobra venom LAAO were tested against several strains of clinical isolates including Staphylococcus aureus, Staphylococcus epidermidis, Pseudomonas aeruginosa, Klebsiella pneumoniae, and Escherichia coli using broth microdilution assay. For comparison, the antibacterial effects of several antibiotics (cefotaxime, kanamycin, tetracycline, vancomycin and penicillin) were also examined using the same conditions. King cobra venom LAAO was very effective in inhibiting the two Gram-positive bacteria (S. aureus and S. epidermidis) tested, with minimum inhibitory concentration (MIC) of 0.78μg/mL (0.006μM) and 1.56μg/mL (0.012μM) against S. aureus and S. epidermidis, respectively. The MICs are comparable to the MICs of the antibiotics tested, on a weight basis. However, the LAAO was only moderately effective against three Gram-negative bacteria tested (P. aeruginosa, K. pneumoniae and E. coli), with MIC ranges from 25 to 50μg/mL (0.2-0.4μM). Catalase at the concentration of 1mg/mL abolished the antibacterial effect of LAAO, indicating that the antibacterial effect of the enzyme involves generation of hydrogen peroxide. Binding studies indicated that king cobra venom LAAO binds strongly to the Gram-positive S. aureus and S. epidermidis, but less strongly to the Gram-negative E. coli and P. aeruginosa, indicating that specific binding to bacteria is important for the potent antibacterial activity of the enzyme. Copyright © 2010 Elsevier Inc. All rights reserved.

Verification of simulation model with COBRA-IIIP code by confrontment of experimental results

International Nuclear Information System (INIS)

Silva Galetti, M.R. da; Pontedeiro, A.C.; Oliveira Barroso, A.C. de

1985-01-01

It is presented an evaluation of the COBRA IIIP/MIT code (of thermal hydraulic analysis by subchannels), comparing their results with experimental data obtained in stationary and transient regimes. It was done a study to calculate the spatial and temporal critical heat flux. It is presented a sensitivity study of simulation model related to the turbulent mixture and the number of axial intervals. (M.C.K.) [pt

Perturbative methods for sensitivity calculation in safety problems of nuclear reactors: state-of-the-art

International Nuclear Information System (INIS)

Lima, Fernando R.A.; Lira, Carlos A.B.O.; Gandini, Augusto

1995-01-01

During the last two decades perturbative methods became an efficient tool to perform sensitivity analysis in nuclear reactor safety problems. In this paper, a comparative study taking into account perturbation formalisms (Diferential and Matricial Mthods and generalized Perturbation Theory - GPT) is considered. Then a few number of applications are described to analyze the sensitivity of some functions relavant to thermal hydraulics designs or safety analysis of nuclear reactor cores and steam generators. The behaviours of the nuclear reactor cores and steam generators are simulated, respectively, by the COBRA-IV-I and GEVAP codes. Results of sensitivity calculations have shown a good agreement when compared to those obtained directly by using the mentioned codes. So, a significative computational time safe can be obtained with perturbative methods performing sensitivity analysis in nuclear power plants. (author). 25 refs., 5 tabs

ENT COBRA (Consortium for Brachytherapy Data Analysis: interdisciplinary standardized data collection system for head and neck patients treated with interventional radiotherapy (brachytherapy

Directory of Open Access Journals (Sweden)

Luca Tagliaferri

2016-08-01

Full Text Available Purpose : Aim of the COBRA (Consortium for Brachytherapy Data Analysis project is to create a multicenter group (consortium and a web-based system for standardized data collection. Material and methods: GEC-ESTRO (Groupe Européen de Curiethérapie – European Society for Radiotherapy & Oncology Head and Neck (H&N Working Group participated in the project and in the implementation of the consortium agreement, the ontology (data-set and the necessary COBRA software services as well as the peer reviewing of the general anatomic site-specific COBRA protocol. The ontology was defined by a multicenter task-group. Results : Eleven centers from 6 countries signed an agreement and the consortium approved the ontology. We identified 3 tiers for the data set: Registry (epidemiology analysis, Procedures (prediction models and DSS, and Research (radiomics. The COBRA-Storage System (C-SS is not time-consuming as, thanks to the use of “brokers”, data can be extracted directly from the single center’s storage systems through a connection with “structured query language database” (SQL-DB, Microsoft Access®, FileMaker Pro®, or Microsoft Excel®. The system is also structured to perform automatic archiving directly from the treatment planning system or afterloading machine. The architecture is based on the concept of “on-purpose data projection”. The C-SS architecture is privacy protecting because it will never make visible data that could identify an individual patient. This C-SS can also benefit from the so called “distributed learning” approaches, in which data never leave the collecting institution, while learning algorithms and proposed predictive models are commonly shared. Conclusions : Setting up a consortium is a feasible and practicable tool in the creation of an international and multi-system data sharing system. COBRA C-SS seems to be well accepted by all involved parties, primarily because it does not influence the center’s own

Effect of Mucuna pruriens Seed Extract Pretreatment on the Responses of Spontaneously Beating Rat Atria and Aortic Ring to Naja sputatrix (Javan Spitting Cobra Venom

Directory of Open Access Journals (Sweden)

Shin Yee Fung

2012-01-01

Full Text Available Mucuna pruriens Linn. (velvet bean has been used by native Nigerians as a prophylactic for snakebite. Rats pretreated with M. pruriens seed extract (MPE have been shown to protect against the lethal and cardiovascular depressant effects of Naja sputatrix (Javan spitting cobra venoms, and the protective effect involved immunological neutralization of the venom toxins. To investigate further the mechanism of the protective effect of MPE pretreatment against cobra venom toxicity, the actions of Naja sputatrix venom on spontaneously beating rat atria and aortic rings isolated from both MPE pretreated and untreated rats were studied. Our results showed that the MPE pretreatment conferred protection against cobra venom-induced depression of atrial contractility and atrial rate in the isolated atrial preparations, but it had no effect on the venom-induced contractile response of aortic ring preparation. These observations suggested that the protective effect of MPE pretreatment against cobra venom toxicity involves a direct protective action of MPE on the heart function, in addition to the known immunological neutralization mechanism, and that the protective effect does not involve action on blood vessel contraction. The results also suggest that M. pruriens seed may contain novel cardioprotective agent with potential therapeutic value.

How the Cobra Got Its Flesh-Eating Venom: Cytotoxicity as a Defensive Innovation and Its Co-Evolution with Hooding, Aposematic Marking, and Spitting

Directory of Open Access Journals (Sweden)

Nadya Panagides

2017-03-01

Full Text Available The cytotoxicity of the venom of 25 species of Old World elapid snake was tested and compared with the morphological and behavioural adaptations of hooding and spitting. We determined that, contrary to previous assumptions, the venoms of spitting species are not consistently more cytotoxic than those of closely related non-spitting species. While this correlation between spitting and non-spitting was found among African cobras, it was not present among Asian cobras. On the other hand, a consistent positive correlation was observed between cytotoxicity and utilisation of the defensive hooding display that cobras are famous for. Hooding and spitting are widely regarded as defensive adaptations, but it has hitherto been uncertain whether cytotoxicity serves a defensive purpose or is somehow useful in prey subjugation. The results of this study suggest that cytotoxicity evolved primarily as a defensive innovation and that it has co-evolved twice alongside hooding behavior: once in the Hemachatus + Naja and again independently in the king cobras (Ophiophagus. There was a significant increase of cytotoxicity in the Asian Naja linked to the evolution of bold aposematic hood markings, reinforcing the link between hooding and the evolution of defensive cytotoxic venoms. In parallel, lineages with increased cytotoxicity but lacking bold hood patterns evolved aposematic markers in the form of high contrast body banding. The results also indicate that, secondary to the evolution of venom rich in cytotoxins, spitting has evolved three times independently: once within the African Naja, once within the Asian Naja, and once in the Hemachatus genus. The evolution of cytotoxic venom thus appears to facilitate the evolution of defensive spitting behaviour. In contrast, a secondary loss of cytotoxicity and reduction of the hood occurred in the water cobra Naja annulata, which possesses streamlined neurotoxic venom similar to that of other aquatic elapid snakes (e

How the Cobra Got Its Flesh-Eating Venom: Cytotoxicity as a Defensive Innovation and Its Co-Evolution with Hooding, Aposematic Marking, and Spitting.

Science.gov (United States)

Panagides, Nadya; Jackson, Timothy N W; Ikonomopoulou, Maria P; Arbuckle, Kevin; Pretzler, Rudolf; Yang, Daryl C; Ali, Syed A; Koludarov, Ivan; Dobson, James; Sanker, Brittany; Asselin, Angelique; Santana, Renan C; Hendrikx, Iwan; van der Ploeg, Harold; Tai-A-Pin, Jeremie; van den Bergh, Romilly; Kerkkamp, Harald M I; Vonk, Freek J; Naude, Arno; Strydom, Morné A; Jacobsz, Louis; Dunstan, Nathan; Jaeger, Marc; Hodgson, Wayne C; Miles, John; Fry, Bryan G

2017-03-13

The cytotoxicity of the venom of 25 species of Old World elapid snake was tested and compared with the morphological and behavioural adaptations of hooding and spitting. We determined that, contrary to previous assumptions, the venoms of spitting species are not consistently more cytotoxic than those of closely related non-spitting species. While this correlation between spitting and non-spitting was found among African cobras, it was not present among Asian cobras. On the other hand, a consistent positive correlation was observed between cytotoxicity and utilisation of the defensive hooding display that cobras are famous for. Hooding and spitting are widely regarded as defensive adaptations, but it has hitherto been uncertain whether cytotoxicity serves a defensive purpose or is somehow useful in prey subjugation. The results of this study suggest that cytotoxicity evolved primarily as a defensive innovation and that it has co-evolved twice alongside hooding behavior: once in the Hemachatus + Naja and again independently in the king cobras ( Ophiophagus ). There was a significant increase of cytotoxicity in the Asian Naja linked to the evolution of bold aposematic hood markings, reinforcing the link between hooding and the evolution of defensive cytotoxic venoms. In parallel, lineages with increased cytotoxicity but lacking bold hood patterns evolved aposematic markers in the form of high contrast body banding. The results also indicate that, secondary to the evolution of venom rich in cytotoxins, spitting has evolved three times independently: once within the African Naja , once within the Asian Naja , and once in the Hemachatus genus. The evolution of cytotoxic venom thus appears to facilitate the evolution of defensive spitting behaviour. In contrast, a secondary loss of cytotoxicity and reduction of the hood occurred in the water cobra Naja annulata , which possesses streamlined neurotoxic venom similar to that of other aquatic elapid snakes (e.g., hydrophiine

The protective effects of Mucuna pruriens seed extract against histopathological changes induced by Malayan cobra (Naja sputatrix) venom in rats.

Science.gov (United States)

Fung, S Y; Tan, N H; Liew, S H; Sim, S M; Aguiyi, J C

2009-04-01

Seed of Mucuna pruriens (Velvet beans) has been prescribed by traditional medicine practitioners in Nigeria as a prophylactic oral antisnake remedy. In the present studies, we investigated the protective effects of M. pruriens seed extract (MPE) against histopathological changes induced by intravenous injection of Naja sputatrix (Malayan cobra) venom in rats pretreated with the seed extract. Examination by light microscope revealed that the venom induced histopathological changes in heart and blood vessels in liver, but no effect on brain, lung, kidney and spleen. The induced changes were prevented by pretreatment of the rats with MPE. Our results suggest that MPE pretreatment protects rat heart and liver blood vessels against cobra venom-induced damages.

Background analysis for the beta-spectrum of the isotope 113Cd in the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Platzek, Stephan [Technische Universitaet Dresden (Germany); Collaboration: COBRA-Collaboration

2016-07-01

The COBRA experiment uses Cadmium-Zinc-Telluride as detector material. This semiconductor contains several isotopes that are candidates for neutrinoless double beta-decay. Due to the natural abundance of the detector material various other isotopes are present as well. One of them is {sup 113}Cd with an abundance of about 12%. The fourfold forbidden non-unique beta-decay of {sup 113}Cd is a rare process with a half-life of about 8.10{sup 15} years. The shape of the spectrum is still topic of scientific discussions because of various forecasts given by theoretical models. The signal related to this decay is by far the most prominent in the COBRA setup causing more than 98% of the total rate. In this talk potential background components contributing to the {sup 113}Cd beta-spectrum are discussed with the aim to develop a detailed background simulation with the program VENOM (based on Geant4), that includes background sources originating from cosmic activation as well as natural radioactivity and detector specific effects.

Substrate recognition by complement convertases revealed in the C5-cobra venom factor complex

DEFF Research Database (Denmark)

Laursen, Nick Stub; Andersen, Kasper Røjkjær; Braren, Ingke

2011-01-01

with a protease subunit (Bb or C2a). We determined the crystal structures of the C3b homologue cobra venom factor (CVF) in complex with C5, and in complex with C5 and the inhibitor SSL7 at 4.3 Å resolution. The structures reveal a parallel two-point attachment between C5 and CVF, where the presence of SSL7 only...

COBRA-LIKE2, a Member of the Glycosylphosphatidylinositol-Anchored COBRA-LIKE Family, Plays a Role in Cellulose Deposition in Arabidopsis Seed Coat Mucilage Secretory Cells1,2[OPEN

Science.gov (United States)

Ben-Tov, Daniela; Abraham, Yael; Stav, Shira; Thompson, Kevin; Loraine, Ann; Elbaum, Rivka; de Souza, Amancio; Pauly, Markus; Kieber, Joseph J.; Harpaz-Saad, Smadar

2015-01-01

Differentiation of the maternally derived seed coat epidermal cells into mucilage secretory cells is a common adaptation in angiosperms. Recent studies identified cellulose as an important component of seed mucilage in various species. Cellulose is deposited as a set of rays that radiate from the seed upon mucilage extrusion, serving to anchor the pectic component of seed mucilage to the seed surface. Using transcriptome data encompassing the course of seed development, we identified COBRA-LIKE2 (COBL2), a member of the glycosylphosphatidylinositol-anchored COBRA-LIKE gene family in Arabidopsis (Arabidopsis thaliana), as coexpressed with other genes involved in cellulose deposition in mucilage secretory cells. Disruption of the COBL2 gene results in substantial reduction in the rays of cellulose present in seed mucilage, along with an increased solubility of the pectic component of the mucilage. Light birefringence demonstrates a substantial decrease in crystalline cellulose deposition into the cellulosic rays of the cobl2 mutants. Moreover, crystalline cellulose deposition into the radial cell walls and the columella appears substantially compromised, as demonstrated by scanning electron microscopy and in situ quantification of light birefringence. Overall, the cobl2 mutants display about 40% reduction in whole-seed crystalline cellulose content compared with the wild type. These data establish that COBL2 plays a role in the deposition of crystalline cellulose into various secondary cell wall structures during seed coat epidermal cell differentiation. PMID:25583925

COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

International Nuclear Information System (INIS)

Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

1986-11-01

COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods

Validation study of the COBRA-WC computer program for LMFBR core thermal-hydraulic analysis

International Nuclear Information System (INIS)

Khan, E.U.; Bates, J.M.

1982-01-01

The COBRA-WC (Whole Core) computer program has been developed as a benchmark code to predict flow and temperature fields in LMFBR rod bundles. Consequently, an extensive validation study has been conducted to reinforce its credibility. A set of generalized parameters predicts data well for a wide range of geometries and operating conditions which include conventional (current generation LMFBRs) fuel and blanket assembly geometry in the forced, mixed, and natural convection regimes. The data base used for validating COBRA-WC was obtained from out-of-pile and in-pile tests. Most of the data was obtained in fully heated bundles with bundle power skew across flats up to 3:1 (max:min), Reynolds number between 500 and 80,000, and coolant mixed-mean temperature rise (δ anti T) in the range, 78 0 F less than or equal to δ anti T less than or equal to 340 0 F. Within the bundle, 95% of the predicted coolant temperature data points fall within +-25 0 F for 150 0 F less than or equal to δ anti T less than or equal to 340 0 F and within +-17 0 F for 78 0 F less than or equal to δ anti T less than or equal to 150 0 F

Studies of Hot Spots in Imploding Wire Arrays at 1 MA on COBRA

International Nuclear Information System (INIS)

Pikuz, Sergey A.; Shelkovenko, Tatiana A.; McBride, Ryan D.; Hammer, David A.

2009-01-01

We present recent results from hot spot investigations in imploding Al wire array z-pinches on the COBRA generator at Cornell University using x-ray diagnostics. Measurements of the temporal and spatial distribution of hot spots in stagnating plasmas by an x-ray streak-camera are included. Experiments show that hot spots have nanosecond lifetime and appear randomly along the array axis after plasma stagnation in secondary pinches in 8 mm diameter and during plasma stagnation in the arrays with 4 mm diameter.

Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

2010-07-01

The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

2016-08-15

Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

The king cobra genome reveals dynamic gene evolution and adaptation in the snake venom system

OpenAIRE

Vonk, Freek J.; Casewell, Nicholas R.; Henkel, Christiaan V.; Heimberg, Alysha M.; Jansen, Hans J.; McCleary, Ryan J. R.; Kerkkamp, Harald M. E.; Vos, Rutger A.; Guerreiro, Isabel; Calvete, Juan J.; Wüster, Wolfgang; Woods, Anthony E.; Logan, Jessica M.; Harrison, Robert A.; Castoe, Todd A.

2013-01-01

Snakes are limbless predators, and many species use venom to help overpower relatively large, agile prey. Snake venoms are complex protein mixtures encoded by several multilocus gene families that function synergistically to cause incapacitation. To examine venom evolution, we sequenced and interrogated the genome of a venomous snake, the king cobra (Ophiophagus hannah), and compared it, together with our unique transcriptome, microRNA, and proteome datasets from this species, with data from ...

Extension of the analytic nodal diffusion solver ANDES to triangular-Z geometry and coupling with COBRA-IIIc for hexagonal core analysis

International Nuclear Information System (INIS)

Lozano, Juan-Andres; Jimenez, Javier; Garcia-Herranz, Nuria; Aragones, Jose-Maria

2010-01-01

In this paper the extension of the multigroup nodal diffusion code ANDES, based on the Analytic Coarse Mesh Finite Difference (ACMFD) method, from Cartesian to hexagonal geometry is presented, as well as its coupling with the thermal-hydraulic (TH) code COBRA-IIIc for hexagonal core analysis. In extending the ACMFD method to hexagonal assemblies, triangular-Z nodes are used. In the radial plane, a direct transverse integration procedure is applied along the three directions that are orthogonal to the triangle interfaces. The triangular nodalization avoids the singularities, that appear when applying transverse integration to hexagonal nodes, and allows the advantage of the mesh subdivision capabilities implicit within that geometry. As for the thermal-hydraulics, the extension of the coupling scheme to hexagonal geometry has been performed with the capability to model the core using either assembly-wise channels (hexagonal mesh) or a higher refinement with six channels per fuel assembly (triangular mesh). Achieving this level of TH mesh refinement with COBRA-IIIc code provides a better estimation of the in-core 3D flow distribution, improving the TH core modelling. The neutronics and thermal-hydraulics coupled code, ANDES/COBRA-IIIc, previously verified in Cartesian geometry core analysis, can also be applied now to full three-dimensional VVER core problems, as well as to other thermal and fast hexagonal core designs. Verification results are provided, corresponding to the different cases of the OECD/NEA-NSC VVER-1000 Coolant Transient Benchmarks.

COBRA - 3C/KFKI: a digital computer program for steady and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

International Nuclear Information System (INIS)

Vigassy, J.; Kovacs, L.M.

1977-11-01

COBRA-3C/KFKI is a digital computer program for the CDC-3300 computer in FORTRAN language. The program is a revised version of the original COBRA-3C code. The code calculates steady-state and transient flow and enthalpy transport in rod-bundle nuclear fuel elements in both boiling and nonboiling conditions. The mathematical model is formulated by dividing the bundle flow area into flow subchannels that are assumed to contain one-dimensional flow and are coupled to each other by turbulent and diversion crossflow mixing. The program neglects sonic velocity propagation but allows for a temporal and spatial acceleration of the diversion crossflow in the transverse momentum equation. A semiexplicit finite-difference scheme is used to perform a boundary-value solution where the boundary conditions are the inlet enthalpy, inlet flow rate and exit pressure. (D.P.)

Refractive index gradient diagnostics: analysis of different optical systems and application to COBRA ion diode

Energy Technology Data Exchange (ETDEWEB)

Knyazev, B A; Greenly, J B; Hammer, D A; Krastelev, E G [Cornell Univ., Ithaca, NY (United States). Laboratory of Plasma Studies; Cuneo, M E [Sandia National Laboratories, Albuquerque, NM (United States)

1997-12-31

Different optical system variations for refractive index gradient diagnostics with a laser beam probe have been analyzed. A `three-telescope` optical system which permits simultaneous measurement of both the laser beam centroid deflection by a bi-cell photodiode and the spatial Fourier spectrum of the deflected beam by a streak camera has been implemented on the COBRA ion diode. The dynamics of the anode plasma layer was studied with these techniques. (author). 3 figs., 8 refs.

COBRA-SFS thermal analysis of a sealed storage cask for the Monitored Retrievable Storage of spent fuel

International Nuclear Information System (INIS)

Rector, D.R.; Wheeler, C.L.

1986-01-01

The COBRA-SFS (Spent Fuel Storage) computer code was used to predict temperature distributions in a concrete Sealed Storage Cask (SSC). This cask was designed for the Department of Energy in the Monitored Retrievable Storage (MRS) program for storage of spent fuel from commercial power operations. Analytical results were obtained for nominal operation of the SSC with spent fuel from 36 PWR fuel assemblies consolidated in 12 cylindrical canisters. Each canister generates 1650 W of thermal power. A parametric study was performed to assess the effects on cask thermal performance of thermal conductivity of the concrete, the fin material, and the amount of radial reinforcing steel bars (rebar). Seven different cases were modeled. The results of the COBRA-SFS analysis of the current cask design predict that the peak fuel cladding temperature in the SSC will not exceed the 37 0 C design limit for the maximum spent fuel load of 19.8 kW and a maximum expected ambient temperature of 37.8 0 C (100 0 F). The results of the parametric analyses illustrate the importance of material selection and design optimization with regard to the SSC thermal performance

Venom Proteomics of Indonesian King Cobra, Ophiophagus hannah: Integrating Top-Down and Bottom-Up Approaches.

Science.gov (United States)

Petras, Daniel; Heiss, Paul; Süssmuth, Roderich D; Calvete, Juan J

2015-06-05

We report on the first application of top-down mass spectrometry in snake venomics. De novo sequence tags generated by, and ProSight Lite supported analysis of, combined collisional based dissotiations (CID and HCD) recorded in a hybrid LTQ Orbitrap instrument in data-dependent mode identified a number of proteins from different toxin families, namely, 11 three-finger toxins (7-7.9 kDa), a Kunitz-type inhibitor (6.3 kDa), ohanin (11.9 kDa), a novel phospholipase A2 molecule (13.8 kDa), and the cysteine-rich secretory protein (CRISP) ophanin (25 kDa) from Indonesian king cobra venom. Complementary bottom-up MS/MS analyses contributed to the completion of a locus-resolved venom phenotypic map for Ophiophagus hannah, the world's longest venomous snake and a species of medical concern across its wide distribution range in forests from India to Southeast Asia. Its venom composition, comprising 32-35 proteins/peptides from 10 protein families, is dominated by α-neurotoxins and convincingly explains the main neurotoxic effects of human envenoming caused by king cobra bite. The integration of efficient chromatographic separation of the venom's components and locus-resolved toxin identification through top-down and bottom-up MS/MS-based species-specific database searching and de novo sequencing holds promise that the future will be bright for the field of venom research.

Thermal-hydraulic modeling of nanofluids as the coolant in VVER-1000 reactor core by the porous media approach

International Nuclear Information System (INIS)

Jahanfarnia, G.; Zarifi, E.; Veysi, F.

2013-01-01

The aim of this study was to perform a thermal-hydraulic analysis of nanofluids as coolant in the Bushehr VVER-1000 reactor core using the porous media approach. Water-based nanofluids containing various volume fractions of Al 2 O 3 and TiO 2 nanoparticles were analyzed. The conservation equations were discretized by the finite volume method and solved by numerical methods. To validate the approaches applied in this study, the results of the model and COBRA-EN code were compared for pure water. The achieved results show that the temperature of the coolant increases with the concentration of the nanoparticles. (authors)

Thermohydraulic analysis of assemblies containing up to 2/7 fuel rods

International Nuclear Information System (INIS)

Ferreira, W.J.; Luz, M.

1985-01-01

The COBRA IV-I computer code was tested using data from the Fast Flux Test Facility. Then this code was applied to the analysis of fuel assemblies from the Binary Breeder Reactor. Previously this analysis was carried out using the COBRA III-C code which allows only for the calculations of fuel assemblies having seven fuel pins. The COBRA IV-I permits the calculation of fuel assemblies containing up to 217 fuel pins and the inclusion of blanket and shielding effects. (F.E.) [pt

Protection set-points lines for the reactor core and considerations about power distribution and peak factors

International Nuclear Information System (INIS)

Furieri, E.B.

1981-01-01

In order to assure the reactor core integrity during the slow operational transients (power excursion above the nominal value and the high coolant temperature), the formation of a steam film (DNB-Departure from Nucleate Boiling) in the control rods must be avoided. The protection set points lines presents the points where DNBR (relation between critical heat flux-q sub(DNB) and the local heat flux-q' sub(local) is equal to 1.30, corrected by peak factors and uncertainty in function of ΔTr and T sub(R), respectively coolant elevation and medium coolant temperature in reactor pressure vessel. The curve set-points were determined using a new version of COBRA-IIIF (CUPRO) computer code, implemented with new subroutines and linearized convergence scheme. Pratical results for Angra-1 core were obtained and its were compared with the results from the fabricator. (E.G.) [pt

Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

International Nuclear Information System (INIS)

Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

2007-12-01

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor

Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

Energy Technology Data Exchange (ETDEWEB)

Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

2007-12-15

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor.

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

2006-01-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

Energy Technology Data Exchange (ETDEWEB)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

2006-07-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

International Nuclear Information System (INIS)

Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

2004-01-01

The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

The king cobra genome reveals dynamic gene evolution and adaptation in the snake venom system.

Science.gov (United States)

Vonk, Freek J; Casewell, Nicholas R; Henkel, Christiaan V; Heimberg, Alysha M; Jansen, Hans J; McCleary, Ryan J R; Kerkkamp, Harald M E; Vos, Rutger A; Guerreiro, Isabel; Calvete, Juan J; Wüster, Wolfgang; Woods, Anthony E; Logan, Jessica M; Harrison, Robert A; Castoe, Todd A; de Koning, A P Jason; Pollock, David D; Yandell, Mark; Calderon, Diego; Renjifo, Camila; Currier, Rachel B; Salgado, David; Pla, Davinia; Sanz, Libia; Hyder, Asad S; Ribeiro, José M C; Arntzen, Jan W; van den Thillart, Guido E E J M; Boetzer, Marten; Pirovano, Walter; Dirks, Ron P; Spaink, Herman P; Duboule, Denis; McGlinn, Edwina; Kini, R Manjunatha; Richardson, Michael K

2013-12-17

Snakes are limbless predators, and many species use venom to help overpower relatively large, agile prey. Snake venoms are complex protein mixtures encoded by several multilocus gene families that function synergistically to cause incapacitation. To examine venom evolution, we sequenced and interrogated the genome of a venomous snake, the king cobra (Ophiophagus hannah), and compared it, together with our unique transcriptome, microRNA, and proteome datasets from this species, with data from other vertebrates. In contrast to the platypus, the only other venomous vertebrate with a sequenced genome, we find that snake toxin genes evolve through several distinct co-option mechanisms and exhibit surprisingly variable levels of gene duplication and directional selection that correlate with their functional importance in prey capture. The enigmatic accessory venom gland shows a very different pattern of toxin gene expression from the main venom gland and seems to have recruited toxin-like lectin genes repeatedly for new nontoxic functions. In addition, tissue-specific microRNA analyses suggested the co-option of core genetic regulatory components of the venom secretory system from a pancreatic origin. Although the king cobra is limbless, we recovered coding sequences for all Hox genes involved in amniote limb development, with the exception of Hoxd12. Our results provide a unique view of the origin and evolution of snake venom and reveal multiple genome-level adaptive responses to natural selection in this complex biological weapon system. More generally, they provide insight into mechanisms of protein evolution under strong selection.

Antimicrobial activity of plants used as medicinals on an indigenous reserve in Rio das Cobras, Paraná, Brazil.

Science.gov (United States)

Moura-Costa, Gislaine F; Nocchi, Samara R; Ceole, Ligia F; de Mello, João Carlos P; Nakamura, Celso Vataru; Dias Filho, Benedito Prado; Temponi, Livia G; Ueda-Nakamura, Tania

2012-09-28

A considerable percentage of global biodiversity is located in Brazil, a country that also has rich cultural and ethnic diversity. In the community of Rio das Cobras, Paraná, plants are still widely used in the health care not only by indigenous people but also by the non-indigenous population that inhabits the region. The investigation of the efficacy and safety of these plants in the treatment of infectious diseases provides insights for future studies of these species allowing the appropriated use by the indigenous people, since few or none study has been conducted so far. Evaluate the antimicrobial activity and cytotoxicity of some plants used as medicinal on an indigenous reserve in Rio das Cobras, Paraná, Brazil. The aqueous extracts were obtained by decoction and the 50% and 70% hydroalcoholic extracts by turbo extraction. The extracts were tested against strains of Staphylococcus aureus, Escherichia coli, Pseudomonas aeruginosa, Bacillus subtilis, Candida albicans, Candida parapsilosis, Candida tropicalis, Leishmania amazonensis, Poliovirus and HSV-1. Cytotoxicity assay using VERO cells were also performed. None of the extracts had a selectivity index (SI)>1 for any of the tested bacteria. Only Campomanesia eugenioides and Schinus terebinthifolius had an SI>1.0 for all of the tested Candida species. The best anti-Leishmania activity was obtained with Zanthoxylum rhoifolium and Schinus terebinthifolius. Extracts of Cordia americana were the most effective against herpes simplex virus type 1. Zanthoxylum rhoifolium was the most effective against Poliovirus, and Ocimum gratissimum was effective against both Poliovirus and Herpes Simplex virus. Among the plants investigated in the present study, Zanthoxylum rhoifolium had the fewest cytotoxic effect. The plants investigated in the present study exhibited potential for future pharmacological uses, but additional studies, especially with regard to in vivo toxicity, must be conducted. The results of this

Exploring the venom of the forest cobra snake: Toxicovenomics and antivenom profiling of Naja melanoleuca

DEFF Research Database (Denmark)

Lauridsen, Line P.; Laustsen, Andreas Hougaard; Lomonte, Bruno

2017-01-01

A toxicovenomic analysis of the venom of the forest cobra, N. melanoleuca, was performed, revealing the presence of a total of 52 proteins by proteomics analysis. The most abundant proteins belong to the three-finger toxins (3FTx) (57.1 wt%), which includes post-synaptically acting α-neurotoxins........ This toxicovenomic study identified the 3FTx group of α-neurotoxins in the venom of N. melanoleuca as the relevant targets to be neutralized....

Determination of the protection set-points lines for the Angra-1 reactor core

International Nuclear Information System (INIS)

Furieri, E.B.

1980-03-01

In this work several thermo-hidraulic calculation were performed to obtain Protection set-points lines for the Angra-1 reactor core in order to compare with the values presented by the vendor in the FSAR. These lines are the locus of points where DNBR min = 1,3 and power = 1,18 x P nominal as a function of ΔT m and T m , the temperature difference and the average coolant temperature between hot and cold legs. A computation scheme was developed using COBRA-IIIF as a subroutine of a new main program and adding new subroutines in order to obtain the desired DNBR. The solution is obtained through a convergentce procedure using parameters estimated in a sensivity study. (author) [pt

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

Comparative Analysis of Thermohydraulic Margins in Embalse Power Station, CARA Vs. CANDU with Cobra IV-HW

International Nuclear Information System (INIS)

Daverio, H; Juanico, L

2000-01-01

Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins

Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

2017-03-15

In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

Photoresponsive nanocapsulation of cobra neurotoxin and enhancement of its central analgesic effects under red light

Directory of Open Access Journals (Sweden)

Yang Q

2017-05-01

Full Text Available Qian Yang, Chuang Zhao, Jun Zhao, Yong Ye Department of Pharmaceutical Engineering, School of Chemistry and Chemical Engineering, South China University of Technology, Guangzhou, People’s Republic of China Abstract: Cobra neurotoxin (CNT, a peptide isolated from snake venom of Naja naja atra, shows central analgesic effects in our previous research. In order to help CNT pass through blood–brain barrier (BBB and improve its central analgesic effects, a new kind of CNT nanocapsules were prepared by double emulsification with soybean lecithin and cholesterol as the shell, and pheophorbide as the photosensitizer added to make it photoresponsive. The analgesic effects were evaluated by hot plate test and acetic acid-induced writhing in mice. The CNT nanocapsules had an average particle size of 229.55 nm, zeta potential of -53.00 mV, encapsulation efficiency of 84.81% and drug loading of 2.98%, when the pheophorbide content was 1% of lecithin weight. Pheophorbide was mainly distributed in outer layer of the CNT nanocapsules and increased the release of the CNT nanocapsules after 650 nm illumination. The central analgesic effects were improved after intraperitoneal injection of CNT at 25 and 50 µg·kg-1 under 650 nm irradiation for 30 min in the nasal cavity. Activation of pheophorbide by red light generated reactive oxygen species which opened the nanocapsules and BBB and helped the CNT enter the brain. This research provides a new drug delivery for treatment of central pain. Keywords: cobra neurotoxin, nanocapsules, photoresponsive, central analgesic effects, red light, drug delivery, photosensitizer

The king cobra genome reveals dynamic gene evolution and adaptation in the snake venom system

Science.gov (United States)

Vonk, Freek J.; Casewell, Nicholas R.; Henkel, Christiaan V.; Heimberg, Alysha M.; Jansen, Hans J.; McCleary, Ryan J. R.; Kerkkamp, Harald M. E.; Vos, Rutger A.; Guerreiro, Isabel; Calvete, Juan J.; Wüster, Wolfgang; Woods, Anthony E.; Logan, Jessica M.; Harrison, Robert A.; Castoe, Todd A.; de Koning, A. P. Jason; Pollock, David D.; Yandell, Mark; Calderon, Diego; Renjifo, Camila; Currier, Rachel B.; Salgado, David; Pla, Davinia; Sanz, Libia; Hyder, Asad S.; Ribeiro, José M. C.; Arntzen, Jan W.; van den Thillart, Guido E. E. J. M.; Boetzer, Marten; Pirovano, Walter; Dirks, Ron P.; Spaink, Herman P.; Duboule, Denis; McGlinn, Edwina; Kini, R. Manjunatha; Richardson, Michael K.

2013-01-01

Snakes are limbless predators, and many species use venom to help overpower relatively large, agile prey. Snake venoms are complex protein mixtures encoded by several multilocus gene families that function synergistically to cause incapacitation. To examine venom evolution, we sequenced and interrogated the genome of a venomous snake, the king cobra (Ophiophagus hannah), and compared it, together with our unique transcriptome, microRNA, and proteome datasets from this species, with data from other vertebrates. In contrast to the platypus, the only other venomous vertebrate with a sequenced genome, we find that snake toxin genes evolve through several distinct co-option mechanisms and exhibit surprisingly variable levels of gene duplication and directional selection that correlate with their functional importance in prey capture. The enigmatic accessory venom gland shows a very different pattern of toxin gene expression from the main venom gland and seems to have recruited toxin-like lectin genes repeatedly for new nontoxic functions. In addition, tissue-specific microRNA analyses suggested the co-option of core genetic regulatory components of the venom secretory system from a pancreatic origin. Although the king cobra is limbless, we recovered coding sequences for all Hox genes involved in amniote limb development, with the exception of Hoxd12. Our results provide a unique view of the origin and evolution of snake venom and reveal multiple genome-level adaptive responses to natural selection in this complex biological weapon system. More generally, they provide insight into mechanisms of protein evolution under strong selection. PMID:24297900

Calculation of mass flow and steam quality distribution on fuel elements of light-water cooled boiling water nuclear reactors

International Nuclear Information System (INIS)

Hermanns, H.J.

1977-04-01

By the example of light-water cooled nuclear reactors, the state of the calculation methods at disposal for calculating mass flow and steam quality distribution (sub-channel analysis) is indicated. Particular regard was paid to the transport phenomena occurring in reactor fuel elements in the range of two phase flow. Experimentally determined values were compared with recalculations of these experiments with the sub-channel code COBRA; from the results of these comparing calculations, conclusions could be drawn on the suitability of this code for defined applications. Limits of reliability could be determined to some extent. Based on the experience gained and the study of individual physical model concepts, recognized as being important, a sub-channel model was drawn up and the corresponding numerical computer code (SIEWAS) worked out. Experiments made at GE could be reproduced with the code SIEWAS with sufficient accuracy. (orig.) [de

Venomics, lethality and neutralization of Naja kaouthia (monocled cobra) venoms from three different geographical regions of Southeast Asia.

Science.gov (United States)

Tan, Kae Yi; Tan, Choo Hock; Fung, Shin Yee; Tan, Nget Hong

2015-04-29

Previous studies showed that venoms of the monocled cobra, Naja kaouthia from Thailand and Malaysia are substantially different in their median lethal doses. The intraspecific venom variations of N. kaouthia, however, have not been fully elucidated. Here we investigated the venom proteomes of N. kaouthia from Malaysia (NK-M), Thailand (NK-T) and Vietnam (NK-V) through reverse-phase HPLC, SDS-PAGE and tandem mass spectrometry. The venom proteins comprise 13 toxin families, with three-finger toxins being the most abundant (63-77%) and the most varied (11-18 isoforms) among the three populations. NK-T has the highest content of neurotoxins (50%, predominantly long neurotoxins), followed by NK-V (29%, predominantly weak neurotoxins and some short neurotoxins), while NK-M has the least (18%, some weak neurotoxins but less short and long neurotoxins). On the other hand, cytotoxins constitute the main bulk of toxins in NK-M and NK-V venoms (up to 45% each), but less in NK-T venom (27%). The three venoms show different lethal potencies that generally reflect the proteomic findings. Despite the proteomic variations, the use of Thai monovalent and Neuro polyvalent antivenoms for N. kaouthia envenomation in the three regions is appropriate as the different venoms were neutralized by the antivenoms albeit at different degrees of effectiveness. Biogeographical variations were observed in the venom proteome of monocled cobra (Naja kaouthia) from Malaysia, Thailand and Vietnam. The Thai N. kaouthia venom is particularly rich in long neurotoxins, while the Malaysian and Vietnamese specimens were predominated with cytotoxins. The differentially expressed toxin profile accounts for the discrepancy in the lethal dose of the venom from different populations. Commercially available Thai antivenoms (monovalent and polyvalent) were able to neutralize the three venoms at different effective doses, hence supporting their uses in the three regions. While dose adjustment according to

Development of realistic thermal hydraulic system analysis code

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

Development of realistic thermal hydraulic system analysis code

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, B. D; Kim, K. D.

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

[A case of favourable outcome of severe acute intoxication with an animal poison after a bite by the monocled cobra].

Science.gov (United States)

Livanov, G A; Batotsyrenkov, B V; Lodiagin, A N; Andrianov, A Iu; Kuznetsov, O A; Loladze, A T; Baranov, D V

2014-01-01

This paper reports a case of severe acute intoxication with an animal poison after a bite by the monocled cobra. Combined treatment including artificial lung ventilation, infusion-detoxication and desensitizing (hormonal) therapy, hemosorption, correction of metabolic disorders with cytoflavin, antibacterial therapy had positive effect on the patient's condition and ensured the favourable outcome ofpotentially lethal poisoning without the use ofa specific anti-snake venom serum.

Perturbative methods applied for sensitive coefficients calculations in thermal-hydraulic systems

International Nuclear Information System (INIS)

Andrade Lima, F.R. de

1993-01-01

The differential formalism and the Generalized Perturbation Theory (GPT) are applied to sensitivity analysis of thermal-hydraulics problems related to pressurized water reactor cores. The equations describing the thermal-hydraulic behavior of these reactors cores, used in COBRA-IV-I code, are conveniently written. The importance function related to the response of interest and the sensitivity coefficient of this response with respect to various selected parameters are obtained by using Differential and Generalized Perturbation Theory. The comparison among the results obtained with the application of these perturbative methods and those obtained directly with the model developed in COBRA-IV-I code shows a very good agreement. (author)

Improving the computation efficiency of COBRA-TF for LWR safety analysis of large problems

International Nuclear Information System (INIS)

Cuervo, D.; Avramova, M. N.; Ivanov, K. N.

2004-01-01

A matrix solver is implemented in COBRA-TF in order to improve the computation efficiency of both numerical solution methods existing in the code, the Gauss elimination and the Gauss-Seidel iterative technique. Both methods are used to solve the system of pressure linear equations and relay on the solution of large sparse matrices. The introduced solver accelerates the solution of these matrices in cases of large number of cells. The execution time is reduced in half as compared to the execution time without using matrix solver for the cases with large matrices. The achieved improvement and the planned future work in this direction are important for performing efficient LWR safety analyses of large problems. (authors)

Substrate recognition by complement convertases revealed in the C5-cobra venom factor complex

DEFF Research Database (Denmark)

Laursen, Nick Stub; Andersen, Kasper Røjkjær; Braren, Ingke

2011-01-01

with a protease subunit (Bb or C2a). We determined the crystal structures of the C3b homologue cobra venom factor (CVF) in complex with C5, and in complex with C5 and the inhibitor SSL7 at 4.3 Å resolution. The structures reveal a parallel two-point attachment between C5 and CVF, where the presence of SSL7 only...... slightly affects the C5-CVF interface, explaining the IgA dependence for SSL7-mediated inhibition of C5 cleavage. CVF functions as a relatively rigid binding scaffold inducing a conformational change in C5, which positions its cleavage site in proximity to the serine protease Bb. A general model...

Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

Energy Technology Data Exchange (ETDEWEB)

Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

2014-07-01

In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

Mapping Proteoforms and Protein Complexes From King Cobra Venom Using Both Denaturing and Native Top-down Proteomics.

Science.gov (United States)

Melani, Rafael D; Skinner, Owen S; Fornelli, Luca; Domont, Gilberto B; Compton, Philip D; Kelleher, Neil L

2016-07-01

Characterizing whole proteins by top-down proteomics avoids a step of inference encountered in the dominant bottom-up methodology when peptides are assembled computationally into proteins for identification. The direct interrogation of whole proteins and protein complexes from the venom of Ophiophagus hannah (king cobra) provides a sharply clarified view of toxin sequence variation, transit peptide cleavage sites and post-translational modifications (PTMs) likely critical for venom lethality. A tube-gel format for electrophoresis (called GELFrEE) and solution isoelectric focusing were used for protein fractionation prior to LC-MS/MS analysis resulting in 131 protein identifications (18 more than bottom-up) and a total of 184 proteoforms characterized from 14 protein toxin families. Operating both GELFrEE and mass spectrometry to preserve non-covalent interactions generated detailed information about two of the largest venom glycoprotein complexes: the homodimeric l-amino acid oxidase (∼130 kDa) and the multichain toxin cobra venom factor (∼147 kDa). The l-amino acid oxidase complex exhibited two clusters of multiproteoform complexes corresponding to the presence of 5 or 6 N-glycans moieties, each consistent with a distribution of N-acetyl hexosamines. Employing top-down proteomics in both native and denaturing modes provides unprecedented characterization of venom proteoforms and their complexes. A precise molecular inventory of venom proteins will propel the study of snake toxin variation and the targeted development of new antivenoms or other biotherapeutics. © 2016 by The American Society for Biochemistry and Molecular Biology, Inc.

Improving the refueling cycle of a WWER-1000 using cuckoo search method and thermal-neutronic coupling of PARCS v2.7, COBRA-EN and WIMSD-5B codes

Energy Technology Data Exchange (ETDEWEB)

Yarizadeh-Beneh, M.; Mazaheri-Beni, H.; Poursalehi, N., E-mail: n_poursalehi@sbu.ac.ir

2016-12-15

Highlights: • The cuckoo search algorithm is applied to the loading pattern optimization of a nuclear reactor core. • Calculations during the cycle show a good agreement between results and reference for the original LP. • Results indicate the efficient performance of cuckoo search approach coupled with thermal-neutronic solvers. • Neutronic parameters of proposed core pattern are improved relative to original core pattern. - Abstract: The fuel loading pattern optimization is an important process in the refueling design of a nuclear reactor core. Also the analysis of reactor core performance during the operation cycle can be a significant step in the core loading pattern optimization (LPO). In this work, for the first time, a new method i.e. cuckoo search algorithm (CS) has been applied to the fuel loading pattern design of Bushehr WWER-1000 core. In this regard, two objectives have been chosen for finding the best configuration including the improvement of operation cycle length associated with flattening the radial power distribution of fuel assemblies. The core pattern optimization has been performed by coupling the CS algorithm to thermal-neutronic codes including PARCS v2.7, COBRA-EN and WIMSD-5B for earning desired parameters along the operation cycle. The calculations have been done for the beginning of cycle (BOC) to the end of cycle (EOC) states. According to numerical results, the longer operation cycle for the semi-optimized loading pattern has been achieved along with less power peaking factor (PPF) in comparison to the original core pattern of Bushehr WWER-1000. Gained results confirm the efficient and suitable performance of the developed program and also the introduced CS method in the LPO of a nuclear WWER type.

CTF Validation and Verification Manual

Energy Technology Data Exchange (ETDEWEB)

Salko, Robert K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Blyth, Taylor S. [Pennsylvania State Univ., University Park, PA (United States); Dances, Christopher A. [Pennsylvania State Univ., University Park, PA (United States); Magedanz, Jeffrey W. [Pennsylvania State Univ., University Park, PA (United States); Jernigan, Caleb [Holtec International, Marlton, NJ (United States); Kelly, Joeseph [U.S. Nuclear Regulatory Commission (NRC), Rockville, MD (United States); Toptan, Aysenur [North Carolina State Univ., Raleigh, NC (United States); Gergar, Marcus [Pennsylvania State Univ., University Park, PA (United States); Gosdin, Chris [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [Pennsylvania State Univ., University Park, PA (United States); Palmtag, Scott [Core Physics, Inc., Cary, NC (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-05-25

Coolant-Boiling in Rod Arrays- Two Fluids (COBRA-TF) is a Thermal/Hydraulic (T/H) simulation code designed for Light Water Reactor (LWR) analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of nine conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and has been used and modified by several institutions over the last several decades. COBRA-TF is also used at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG), and has been improved, updated, and subsequently became the PSU RDFMG version of COBRA-TF (CTF). One part of the improvement process includes validating the methods in CTF. This document seeks to provide a certain level of certainty and confidence in the predictive capabilities of the code for the scenarios it was designed to model--rod bundle geometries with operating conditions that are representative of prototypical Pressurized Water Reactor (PWR)s and Boiling Water Reactor (BWR)s in both normal and accident conditions. This is done by modeling a variety of experiments that simulate these scenarios and then presenting a qualitative and quantitative analysis of the results that demonstrates the accuracy to which CTF is capable of capturing specific quantities of interest.

MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

Exploring the venom of the forest cobra snake: Toxicovenomics and antivenom profiling of Naja melanoleuca.

Science.gov (United States)

Lauridsen, Line P; Laustsen, Andreas H; Lomonte, Bruno; Gutiérrez, José María

2017-01-06

A toxicovenomic analysis of the venom of the forest cobra, N. melanoleuca, was performed, revealing the presence of a total of 52 proteins by proteomics analysis. The most abundant proteins belong to the three-finger toxins (3FTx) (57.1wt%), which includes post-synaptically acting α-neurotoxins. Phospholipases A 2 (PLA 2 ) were the second most abundant group of proteins (12.9wt%), followed by metalloproteinases (SVMPs) (9.7wt%), cysteine-rich secretory proteins (CRISPs) (7.6wt%), and Kunitz-type serine proteinase inhibitors (3.8wt%). A number of additional protein families comprised each <3wt% of venom proteins. A toxicity screening of the fractions, using the mouse lethality test, identified toxicity in RP-HPLC peaks 3, 4, 5 and 8, all of them containing α-neurotoxins of the 3FTx family, whereas the rest of the fractions did not show toxicity at a dose of 0.53mg/kg. Three polyspecific antivenoms manufactured in South Africa and India were tested for their immunoreactivity against crude venom and fractions of N. melanoleuca. Overall, antivenoms immunorecognized all fractions in the venom, the South African antivenom showing a higher titer against the neurotoxin-containing fractions. This toxicovenomic study identified the 3FTx group of α-neurotoxins in the venom of N. melanoleuca as the relevant targets to be neutralized. A toxicovenomic analysis of the venom of the forest cobra, also known as black cobra, Naja melanoleuca, was performed. Envenomings by this elapid species are characterized by a progressive descending paralysis which starts with palpebral ptosis and, in severe cases, ends up with respiratory arrest and death. A total of 52 different proteins were identified in this venom. The most abundant protein family was the three-finger toxin (3FTx) family, which comprises almost 57.1wt% of the venom, followed by phospholipases A 2 (PLA 2 ) (12.9wt%). In addition, several other protein families were identified in a much lower percentage in the venom. A

Viper and cobra venom neutralization by beta-sitosterol and stigmasterol isolated from the root extract of Pluchea indica Less. (Asteraceae).

Science.gov (United States)

Gomes, A; Saha, Archita; Chatterjee, Ipshita; Chakravarty, A K

2007-09-01

We reported previously that the methanolic root extract of the Indian medicinal plant Pluchea indica Less. (Asteraceae) could neutralize viper venom-induced action [Alam, M.I., Auddy, B., Gomes, A., 1996. Viper venom neutralization by Indian medicinal plant (Hemidesmus indicus and P. indica) root extracts. Phytother. Res. 10, 58-61]. The present study reports the neutralization of viper and cobra venom by beta-sitosterol and stigmasterol isolated from the root extract of P. indica Less. (Asteraceae). The active fraction (containing the major compound beta-sitosterol and the minor compound stigmasterol) was isolated and purified by silica gel column chromatography and the structure was determined using spectroscopic analysis (EIMS, (1)H NMR, (13)C NMR). Anti-snake venom activity was studied in experimental animals. The active fraction was found to significantly neutralize viper venom-induced lethal, hemorrhagic, defibrinogenation, edema and PLA(2) activity. Cobra venom-induced lethality, cardiotoxicity, neurotoxicity, respiratory changes and PLA(2) activity were also antagonized by the active component. It potentiated commercial snake venom antiserum action against venom-induced lethality in male albino mice. The active fraction could antagonize venom-induced changes in lipid peroxidation and superoxide dismutase activity. This study suggests that beta-sitosterol and stigmasterol may play an important role, along with antiserum, in neutralizing snake venom-induced actions.

Assessment of 4x4 rod bundle subchannel mixing experiments

International Nuclear Information System (INIS)

Otero, Fatima; Veloso, Maria A.; Pereira, Claubia; Fortini, Angela; Lombardi, Antonella

2011-01-01

An assessment of mixing data taking from a 4x4 rod bundle array, under operating conditions typical of a Boiling Water Reactor (BWR), conducted at Columbia University Heat Transfer Research Facility has been accomplished by using the STHIRP-1 code, which is a UFMG version of the COBRA-3C subchannel code. Although designed for subchannel analysis of research reactor cores, all the capability of COBRA-3C has been preserved in the STHIRP-1 code. In the light of alternative models for turbulent mixing, steam quality, and void fraction, results predicted by this code will be compared with experimental data for specific enthalpy and mass flow rate measured at the exit of two specific subchannels.(author)

The development and application of a sub-channel code in ocean environment

International Nuclear Information System (INIS)

Wu, Pan; Shan, Jianqiang; Xiang, Xiong; Zhang, Bo; Gou, Junli; Zhang, Bin

2016-01-01

Highlights: • A sub-channel code named ATHAS/OE is developed for nuclear reactors in ocean environment. • ATHAS/OE is verified by another modified sub-channel code based on COBRA-IV. • ATHAS/OE is used to analyze thermal hydraulic of a typical SMR in heaving and rolling motion. • Calculation results show that ocean condition affect the thermal hydraulic of a reactor significantly. - Abstract: An upgraded version of ATHAS sub-channel code ATHAS/OE is developed for the investigation of the thermal hydraulic behavior of nuclear reactor core in ocean environment with consideration of heaving and rolling motion effect. The code is verified by another modified sub-channel code based on COBRA-IV and used to analyze the thermal hydraulic characteristics of a typical SMR under heaving and rolling motion condition. The calculation results show that the heaving and rolling motion affect the thermal hydraulic behavior of a reactor significantly.

Inhibition of vitamin D2-induced arteriosclerosis in rats by depletion of complement with cobra venom factor.

Science.gov (United States)

Pang, A S; Minta, J O

1980-01-01

Widespread calcerous deposits developed in the aorta, heart and kidneys of rats fed for 4 days with purina chow and high doses of vitamin D2 (200,000 IU/kg body wt/day). Decomplementation of rats with highly purified cobra venom factor (CoF) prior to vitamin D2 feeding, almost completely prevented calcium deposition in the aorta and arteritis. The mortality rate in the CoF-treated vitamin D2-fed rats was much lower than in untreated rats. These findings suggest that the complement system may be recruited in the pathogenesis of vitamin D2-induced arteriosclerosis.

Culturable Aerobic and Facultative Anaerobic Intestinal Bacterial Flora of Black Cobra (Naja naja karachiensis) in Southern Pakistan

Science.gov (United States)

Iqbal, Junaid; Sagheer, Mehwish; Tabassum, Nazneen; Siddiqui, Ruqaiyyah; Khan, Naveed Ahmed

2014-01-01

Using morphological analysis and biochemical testing, here for the first time, we determined the culturable gut bacterial flora (aerobes and facultative anaerobes) in the venomous Black Cobra (Naja naja karachiensis) from South Asia. The findings revealed that these snakes inhabit potentially pathogenic bacteria including Serratia marcescens, Pseudomonas aeruginosa, Shewanella putrefaciens, Aeromonas hydrophila, Salmonella sp., Moraxella sp., Bacillus sp., Ochrobactrum anthropi, and Providencia rettgeri. These findings are of concern, as injury from snake bite can result in wound infections and tissue necrosis leading to sepsis/necrotizing fasciitis and/or expose consumers of snake meat/medicine in the community to infections. PMID:25002979

Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2014-01-01

The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that - beyond the uncertainties of the geometry and the boundary conditions - it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.

Safety and efficacy of low-dose paclitaxel utilizing the cobra-P drug-eluting stent system with a novel biodegradable coating in de novo coronary lesions: The PLUS-ONE first-in-man study

International Nuclear Information System (INIS)

Calderas, Carlos; Condado, Jose Francisco; Condado, Jose Antonio; Flores, Alejandra; Mueller, Amy; Thomas, Jack; Nakatani, Daisaku; Honda, Yasuhiro; Waseda, Katsuhisa; Fitzgerald, Peter

2014-01-01

Background: The Cobra-P drug-eluting stent (DES) system consists of cobalt chromium alloy with bio-absorbable siloxane sol–gel matrix coating that elutes low dose paclitaxel within 6 months. The aim of this first-in-man trial was to evaluate the safety and performance of 2 doses of the Cobra-P DES. Methods: A total of 60 lesions (54 patients) were sequentially assigned to 2 different paclitaxel doses: group A (3.7 μg/18 mm, n = 30) or group B (8 μg/18 mm, n = 30). The primary endpoint was MACE at 4 months defined as cardiac death, myocardial infarction, and target lesion revascularization. Results: Patient and lesion characteristics were matched between the 2 groups except for male sex. MACE at 4 months was 3.3% and 0% respectively (P = 1.000) and at 1-year follow-up remained unchanged. In-stent late loss at 4 months was similar in both groups (0.36 ± 0.30 mm and 0.34 ± 0.20 mm P = .773). Conclusions: In this FIM study, implantation of the Cobra-P low dose paclitaxel-eluting stent with a bioabsorbable sol–gel coating was proven to be feasible and safe. Moderate neointimal proliferation was observed as well as an acceptable MACE rate up to 1 year

Safety and efficacy of low-dose paclitaxel utilizing the cobra-P drug-eluting stent system with a novel biodegradable coating in de novo coronary lesions: the PLUS-ONE first-in-man study.

Science.gov (United States)

Calderas, Carlos; Condado, Jose Francisco; Condado, Jose Antonio; Flores, Alejandra; Mueller, Amy; Thomas, Jack; Nakatani, Daisaku; Honda, Yasuhiro; Waseda, Katsuhisa; Fitzgerald, Peter

2014-01-01

The Cobra-P drug-eluting stent (DES) system consists of cobalt chromium alloy with bio-absorbable siloxane sol-gel matrix coating that elutes low dose paclitaxel within 6 months. The aim of this first-in-man trial was to evaluate the safety and performance of 2 doses of the Cobra-P DES. A total of 60 lesions (54 patients) were sequentially assigned to 2 different paclitaxel doses: group A (3.7 μg/18mm, n=30) or group B (8 μg/18mm, n=30). The primary endpoint was MACE at 4 months defined as cardiac death, myocardial infarction, and target lesion revascularization. Patient and lesion characteristics were matched between the 2 groups except for male sex. MACE at 4 months was 3.3% and 0% respectively (P=1.000) and at 1-year follow-up remained unchanged. In-stent late loss at 4 months was similar in both groups (0.36 ± 0.30mm and 0.34 ± 0.20mm P=.773). In this FIM study, implantation of the Cobra-P low dose paclitaxel-eluting stent with a bioabsorbable sol-gel coating was proven to be feasible and safe. Moderate neointimal proliferation was observed as well as an acceptable MACE rate up to 1 year. © 2014.

Safety and efficacy of low-dose paclitaxel utilizing the cobra-P drug-eluting stent system with a novel biodegradable coating in de novo coronary lesions: The PLUS-ONE first-in-man study

Energy Technology Data Exchange (ETDEWEB)

Calderas, Carlos [Instituto de Clinicas Urologia Tamanaco, Caracas (Venezuela, Bolivarian Republic of); Condado, Jose Francisco; Condado, Jose Antonio [Hospital Centro Medico de Caracas y Hospital Miguel Perez Carreno, Caracas (Venezuela, Bolivarian Republic of); Flores, Alejandra [Instituto de Clinicas Urologia Tamanaco, Caracas (Venezuela, Bolivarian Republic of); Mueller, Amy; Thomas, Jack [Medlogics Device Corporation, Santa Rosa, CA (United States); Nakatani, Daisaku; Honda, Yasuhiro; Waseda, Katsuhisa [Stanford University, Stanford, CA (United States); Fitzgerald, Peter, E-mail: crci-cvmed@stanford.edu [Stanford University, Stanford, CA (United States)

2014-01-15

Background: The Cobra-P drug-eluting stent (DES) system consists of cobalt chromium alloy with bio-absorbable siloxane sol–gel matrix coating that elutes low dose paclitaxel within 6 months. The aim of this first-in-man trial was to evaluate the safety and performance of 2 doses of the Cobra-P DES. Methods: A total of 60 lesions (54 patients) were sequentially assigned to 2 different paclitaxel doses: group A (3.7 μg/18 mm, n = 30) or group B (8 μg/18 mm, n = 30). The primary endpoint was MACE at 4 months defined as cardiac death, myocardial infarction, and target lesion revascularization. Results: Patient and lesion characteristics were matched between the 2 groups except for male sex. MACE at 4 months was 3.3% and 0% respectively (P = 1.000) and at 1-year follow-up remained unchanged. In-stent late loss at 4 months was similar in both groups (0.36 ± 0.30 mm and 0.34 ± 0.20 mm P = .773). Conclusions: In this FIM study, implantation of the Cobra-P low dose paclitaxel-eluting stent with a bioabsorbable sol–gel coating was proven to be feasible and safe. Moderate neointimal proliferation was observed as well as an acceptable MACE rate up to 1 year.

On the extension of the analytic nodal diffusion solver ANDES to sodium fast reactors

International Nuclear Information System (INIS)

Ochoa, R.; Herrero, J.J.; Garcia-Herranz, N.

2011-01-01

Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor physics group at UPM is working on the extension of its in-house multi-scale advanced deterministic code COBAYA3 to Sodium Fast Reactors (SFR). COBAYA3 is a 3D multigroup neutron kinetics diffusion code that can be used either as a pin-by-pin code or as a stand-alone nodal code by using the analytic nodal diffusion solver ANDES. It is coupled with thermal-hydraulics codes such as COBRA-TF and FLICA, allowing transient analysis of LWR at both fine-mesh and coarse-mesh scales. In order to enable also 3D pin-by-pin and nodal coupled NK-TH simulations of SFR, different developments are in progress. This paper presents the first steps towards the application of COBAYA3 to this type of reactors. ANDES solver, already extended to triangular-Z geometry, has been applied to fast reactor steady-state calculations. The required cross section libraries were generated with ERANOS code for several configurations. Here some of the limitations encountered when attempting to apply the Analytical Coarse Mesh Finite Difference (ACMFD) method - implemented inside ANDES - to fast reactor calculations are discussed and the sensitivity of the method to the energy-group structure is studied. In order to reinforce some of the conclusions obtained two calculations are presented. The first one involves a 3D mini-core model in 33 groups, where the ANDES solver presents several issues. And secondly, a benchmark from the NEA for a small 3D FBR in hexagonal-Z geometry in 4 energy groups is used to verify the good convergence of the code in a few-energy-group structure. (author)

Prediction of flow recirculation in a blanket assembly under worst-case natural-convection conditions

International Nuclear Information System (INIS)

Khan, E.U.; Rector, D.R.

1982-01-01

Reactor fuel and blanket assemblies within a Liquid Metal Fast Breeder Reactor (LMFBR) can be subjected to severe radial heat flux gradients. At low-flow conditions, with power-to-flow ratios of nearly the same magnitude as design conditions, buoyancy forces cause flow redistribution to the side of a bundle with the higher heat generation rate. Recirculation of fluid within a rod bundle can occur during a natural convection transient because of the combined effect of flow coastdown and buoyancy-induced redistribution. An important concern is whether recirculation leads to high coolant temperatures. For this reason, the COBRA-WC code was developed with the capability of modeling recirculating flows. Experiments have been conducted in a 2 x 6 rod bundle for flow and power transients to study recirculation in the mixed-convection (forced cooled) and natural-convection regimes. The data base developed was used to validate the recirculation module in the COBRA-WC code. COBRA-WC code calculations were made to predict flow and temperature distributions in a typical LMFBR blanket assembly for the worst-case, natural-circulation transient

Anticancer Activity of Cobra Venom Polypeptide, Cytotoxin-II, against Human Breast Adenocarcinoma Cell Line (MCF-7) via the Induction of Apoptosis

OpenAIRE

Ebrahim, Karim; Shirazi, Farshad H.; Vatanpour, Hosein; zare, Abas; Kobarfard, Farzad; Rabiei, Hadi

2014-01-01

Purpose Breast cancer is a significant health problem worldwide, accounting for a quarter of all cancer diagnoses in women. Current strategies for breast cancer treatment are not fully effective, and there is substantial interest in the identification of novel anticancer agents especially from natural products including toxins. Cytotoxins are polypeptides found in the venom of cobras and have various physiological effects. In the present study, the anticancer potential of cytotoxin-II against...

In Vitro and In Vivo Evaluation of Polyherbal Formulation against Russell's Viper and Cobra Venom and Screening of Bioactive Components by Docking Studies

Science.gov (United States)

Sakthivel, G.; Dey, Amitabha; Nongalleima, Kh.; Chavali, Murthy; Rimal Isaac, R. S.; Singh, N. Surjit; Deb, Lokesh

2013-01-01

The present study emphasizes to reveal the antivenom activity of Aristolochia bracteolata Lam., Tylophora indica (Burm.f.) Merrill, and Leucas aspera S. which were evaluated against venoms of Daboia russelli russelli (Russell's viper) and Naja naja (Indian cobra). The aqueous extracts of leaves and roots of the above-mentioned plants and their polyherbal (1 : 1 : 1) formulation at a dose of 200 mg/kg showed protection against envenomed mice with LD50 doses of 0.44 mg/kg and 0.28 mg/kg against Russell's viper and cobra venom, respectively. In in vitro antioxidant activities sample extracts showed free radical scavenging effects in dose dependent manner. Computational drug design and docking studies were carried out to predict the neutralizing principles of type I phospholipase A2 (PLA2) from Indian common krait venom. This confirmed that aristolochic acid and leucasin can neutralize type I PLA2 enzyme. Results suggest that these plants could serve as a source of natural antioxidants and common antidote for snake bite. However, further studies are needed to identify the lead molecule responsible for antidote activity. PMID:23533518

Study of rare neutron induced processes and coincidence analyses to identify and reduce background contributions in the COBRA experiment

International Nuclear Information System (INIS)

Timm, Jan Horst Karl

2015-11-01

The aim of the COBRA experiment is the observation of neutrinoless double-beta decay, primarily of the isotope 116 Cd. The applied semiconductor detectors of cadmium zinc telluride that are 90% to be enriched enable both the detection and the source of this decay. The half-lives of decays of this kind are expected in the range of more than 10 26 years. Therefore, the reduction of contributions to the background is of decisive importance. The main subjects of this work are, on the one hand, the time synchronization of the data, which provides the basis for coincidence analysis. This analysis method has access not only to identification of contributions to the background, but also to observe decays involving positron annihilation and decays into excited states. In this study, the intrinsic detector contamination of some decay products of 238 U and 232 Th was measured and sensitivities to the half-lives of the decays like 120 Te and 128 Te in each case to the first excited state of daughter products are given. On the other hand, qualitative studies on the importance of neutrons in the COBRA experiment were conducted. These have shown that fast neutrons, thus with energies greater than 10 keV, only result in an insignificant contribution to the background for the detection of neutrinoless double-beta decay of the 116 Cd. Previous studies have also shown that the thermal neutron flux can be in situ determined by coincidence analysis.

Pharmacokinetics of Naja sumatrana (equatorial spitting cobra venom and its major toxins in experimentally envenomed rabbits.

Directory of Open Access Journals (Sweden)

Michelle Khai Khun Yap

2014-06-01

Full Text Available The optimization of snakebite management and the use of antivenom depend greatly on the knowledge of the venom's composition as well as its pharmacokinetics. To date, however, pharmacokinetic reports on cobra venoms and their toxins are still relatively limited. In the present study, we investigated the pharmacokinetics of Naja sumatrana (Equatorial spitting cobra venom and its major toxins (phospholipase A2, neurotoxin and cardiotoxin, following intravenous and intramuscular administration into rabbits.The serum antigen concentration-time profile of the N. sumatrana venom and its major toxins injected intravenously fitted a two-compartment model of pharmacokinetics. The systemic clearance (91.3 ml/h, terminal phase half-life (13.6 h and systemic bioavailability (41.9% of N. sumatrana venom injected intramuscularly were similar to those of N. sputatrix venom determined in an earlier study. The venom neurotoxin and cardiotoxin reached their peak concentrations within 30 min following intramuscular injection, relatively faster than the phospholipase A2 and whole venom (Tmax=2 h and 1 h, respectively. Rapid absorption of the neurotoxin and cardiotoxin from the injection site into systemic circulation indicates fast onsets of action of these principal toxins that are responsible for the early systemic manifestation of envenoming. The more prominent role of the neurotoxin in N. sumatrana systemic envenoming is further supported by its significantly higher intramuscular bioavailability (Fi.m.=81.5% compared to that of the phospholipase A2 (Fi.m.=68.6% or cardiotoxin (Fi.m.=45.6%. The incomplete absorption of the phospholipase A2 and cardiotoxin may infer the toxins' affinities for tissues at the injection site and their pathological roles in local tissue damages through synergistic interactions.Our results suggest that the venom neurotoxin is absorbed very rapidly and has the highest bioavailability following intramuscular injection, supporting its

Master-3.0: multi-purpose analyzer for static and transient effects of reactors

International Nuclear Information System (INIS)

Cho, Byung Oh; Joo, Han Gyu; Cho, Jin Young; Song, Jae Seung; Zee, Sung Quun

2002-03-01

MASTER-3.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the multi-group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM (Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with NTPEN (Non-linear Triangle-based Polynomial Expansion Nodal Method), AFEN (Analytic Function Expansion Nodal)/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method, energy group restriction/prolongation method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. MASTER-3.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P or MATRA model can be used selectively. In addition, MASTER-3.0 is designed to cover various PWRs including SMART as well as WH- and CE-type reactors, providing all data required in their design procedures

Exploration of Pixelated detectors for double beta decay searches within the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Schwenke, M., E-mail: schwenke@asp.tu-dresden.de [Institut fuer Kern- und Teilchenphysik, Technische Universitaet Dresden, Zellescher Weg 19, 01069 Dresden (Germany); Zuber, K.; Janutta, B. [Institut fuer Kern- und Teilchenphysik, Technische Universitaet Dresden, Zellescher Weg 19, 01069 Dresden (Germany); He, Z.; Zeng, F. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan 48109-2104 (United States); Anton, G.; Michel, T.; Durst, J.; Lueck, F.; Gleixner, T. [Erlangen Centre for Astroparticle Physics, Friedrich-Alexander-Universitaet Erlangen-Nuernberg, Erwin-Rommel-Str. 1, 91058 Erlangen (Germany); Goessling, C.; Schulz, O.; Koettig, T. [Technische Universitaet Dortmund, Physik E IV, 44221 Dortmund (Germany); Krawczynski, H.; Martin, J. [Department of Physics, Washington University in St. Louis, Campus Box 1105, One Brookings Drive, St. Louis, MO 63130-4899 (United States); Stekl, I.; Cermak, P. [Institute of Experimental and Applied Physics, Czech Technical University in Prague, Horska 3a/22, 128 00 Prague (Czech Republic)

2011-09-11

The aim of the COBRA experiment is the search for neutrinoless double beta decay events in Cadmium Zinc Telluride (CdZnTe) room temperature semiconductor detectors. The development of pixelated detectors provides the potential for clear event identification and thus major background reduction. The tracking option of a semiconductor is a unique approach in this field. For initial studies, several possible detector systems are considered with a special regard for low background applications: the large volume system Polaris with a pixelated CdZnTe sensor, Timepix detectors with Si and enriched CdTe sensor material and a CdZnTe pixel system developed at the Washington University in St. Louis, USA. For all detector systems first experimental background measurements taken at underground laboratories (Gran Sasso Underground Laboratory in Italy, LNGS and the Niederniveau Messlabor Felsenkeller in Dresden, Germany) and additionally for the Timepix detectors simulation results are presented.

Benchmark Simulation for the Development of the Regulatory Audit Subchannel Analysis Code

Energy Technology Data Exchange (ETDEWEB)

Lee, G. H.; Song, C.; Woo, S. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-05-15

For the safe and reliable operation of a reactor, it is important to predict accurately the flow and temperature distributions in the thermal-hydraulic design of a reactor core. A subchannel approach can give the reasonable flow and temperature distributions with the short computing time. Korea Institute of Nuclear Safety (KINS) is presently reviewing new subchannel code, THALES, which will substitute for both THINC-IV and TORC code. To assess the prediction performance of THALES, KINS is developing the subchannel analysis code for the independent audit calculation. The code is based on workstation version of COBRA-IV-I. The main objective of the present study is to assess the performance of COBRA-IV-I code by comparing the simulation results with experimental ones for the sample problems

The problems of calculation of heat transfer crisis in fuel assemblies of PW reactors based on modern versions of thermohydraulic codes

International Nuclear Information System (INIS)

Fialko, N.M.; Sharaevskij, G.I.; Sharaevskaya, E.I.; Babak, E.I.

2014-01-01

The article gives an analysis of the adequacy of computer software systems FASCICLE BM-DF and COBRA, which are designed to calculate the main parameters of the safety of water-cooled nuclear reactors. This calculation is based on determining the local thermal-hydraulic parameters of the flow of coolant in the fuel rod assembled elements. In this article introduced the results of the comparison of experiments performed to determine the distribution of the main thermal-hydraulic flow parameters characteristic of subchannels of fuel rod assembled elements with the data for calculating these parameters on the basis of declared computer codes. Particular attention is paid to the analysis of experimental and calculation data, by definition, burnout in rod fuel assembled elements

Development of a shield based on Monte-Carlo studies for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Heidrich, Nadine [Institut fuer Experimentalphysik, 22761 Hamburg (Germany); Collaboration: COBRA-Collaboration

2013-07-01

COBRA is a next-generation experiment searching for neutrinoless double beta decay using CdZnTe semiconductor detectors. The main focus is on {sup 116}Cd, with a decay energy of 2813.5 keV well above the highest naturally occurring gamma lines. The concept for a large scale set-up consists of an array of CdZnTe detectors with a total mass of 420 kg enriched in {sup 116}Cd up to 90 %. With a background rate in the order of 10{sup -3} counts/keV/kg/year, the experiment would be sensitive to a half-life larger than 10{sup 26} years, corresponding to a Majorana mass term m{sub ββ} smaller than 50 meV. To achieve the background level, an appropriate shield is necessary. The shield is developed based on Monte-Carlo simulations. For that, different materials and configurations are tested. In the talk the current status of the Monte-Carlo survey is presented and discussed.

From Cobra Grubs to Dragons: Negotiating the Politics of Representation in Cultural Research

Directory of Open Access Journals (Sweden)

Tanja Dreher

2011-11-01

Full Text Available ‘From cobra grubs to dragons’ was suggested as the title for a cultural tour of the Fairfield area in Sydney developed by the author and others through a partnership between the Centre for Cultural Research (CCR at the University of Western Sydney (UWS, the Fairfield City Council (FCC and the Migration Heritage Centre of the NSW government. The cultural researchers involved in producing the tour felt that this title was an evocative description of the tour which enables participants to visit numerous sites associated with Fairfield’s cultural diversity. As the project developed, a key funding partner vetoed that title, to be replaced by the prosaic ‘Tune in to Fairfield City: a multicultural driving tour’ which now appears on the audio CD and map which contain instructions for motorists to take the tour. This article analyses the rather more complex issues of naming and representation, voice and authority raised by this cultural research project.

Testing and COBRA-SFS analysis of the VSC-17 ventilated concrete, spent fuel storage cask

International Nuclear Information System (INIS)

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-04-01

A performance test of a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask loaded with 17 canisters of consolidated PWR spent fuel generating approximately 15 kW was conducted. The performance test included measuring the cask surface, concrete, air channel surface, and fuel temperatures, as well as cask surface gamma and neutron dose rates. Testing was performed using vacuum, nitrogen, and helium backfill environments. Pretest predictions of cask thermal performance were made using the COBRA-SFS computer code. Analysis results were within 15 degrees C of measured peak fuel temperature. Peak fuel temperature for normal operation was 321 degrees C. In general, the surface dose rates were less than 30 mrem/h on the side of the cask and 40 mrem/h on the top of the cask

Thermal - hydraulic analysis of pressurizer water reactors using the model of open lateral boundary

International Nuclear Information System (INIS)

Borges, R.C.

1980-10-01

A computational method is developed for thermal-hydraulic analysis, where the channel may be analysed by more than one independent steps of calculation. This is made possible by the incorporation of the model of open lateral boundary in the code COBRA-IIIP, which permits the determination of the subchannel of an open lattice PWR core in a multi-step calculation. The thermal-hydraulic code COBRA-IIIP, developed at the Massachusetts Institute of Technology, is used as the basic model for this study. (Author) [pt

Inhibition of the nicotinic acetylcholine receptors by cobra venom α-neurotoxins: is there a perspective in lung cancer treatment?

Directory of Open Access Journals (Sweden)

Angela Alama

Full Text Available Nicotine exerts its oncogenic effects through the binding to nicotinic acetylcholine receptors (nAChRs and the activation of downstream pathways that block apoptosis and promote neo-angiogenesis. The nAChRs of the α7 subtype are present on a wide variety of cancer cells and their inhibition by cobra venom neurotoxins has been proposed in several articles and reviews as a potential innovative lung cancer therapy. However, since part of the published results was recently retracted, we believe that the antitumoral activity of cobra venom neurotoxins needs to be independently re-evaluated.We determined the activity of α-neurotoxins from Naja atra (short-chain neurotoxin, α-cobrotoxin and Naja kaouthia (long-chain neurotoxin, α-cobratoxin in vitro by cytotoxicity measurements in 5 lung cancer cell lines, by colony formation assay with α7nAChRs expressing and non-expressing cell lines and in vivo by assessing tumor growth in an orthotopic Non-Obese Diabetic/Severe Combined Immunodeficient (NOD/SCID mouse model system utilizing different treatment schedules and dosages.No statistically significant reduction in tumor growth was observed in the treatment arms in comparison to the control for both toxins. Paradoxically α-cobrotoxin from Naja atra showed the tendency to enhance tumor growth although, even in this case, the statistical significance was not reached.In conclusion our results show that, in contrast with other reports, the nAChR inhibitors α-cobratoxin from N. kaouthia and α-cobrotoxin from N. atra neither suppressed tumor growth nor prolonged the survival of the treated animals.

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

International Nuclear Information System (INIS)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-01

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

Selection and benchmarking of computer codes for research reactor core conversions

International Nuclear Information System (INIS)

Yilmaz, E.; Jones, B.G.

1983-01-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC 2 , COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of Illinois. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general

Development and assessment of a sub-channel code applicable for trans-critical transient of SCWR

International Nuclear Information System (INIS)

Liu, X.J.; Yang, T.; Cheng, X.

2013-01-01

Highlights: • A new sub-channel code COBRA-SC for SCWR is developed. • Pseudo two-phase method is employed to realize trans-critical transient calculation. • Good suitability of COBRA-SC is demonstrated by preliminary assessment. • The calculation results of COBRA-SC agree well with ATHLET code. -- Abstract: In the last few years, extensive R and D activities have been launched covering various aspects of supercritical water-cooled reactor (SCWR), especially the thermal-hydraulic analysis. Sub-channel code plays an indispensable role to predict the detail thermal-hydraulic behavior of the SCWR fuel assembly. This paper develops a new version of sub-channel code COBRA-SC based on the previous COBRA-IV code. The supercritical water property and heat transfer/pressure drop correlations under supercritical pressure are implemented to this code. Moreover, in order to simulate the trans-critical transient (the pressure undergo a decrease from the supercritical pressure to the subcritical pressure), pseudo two-phase method is employed in COBRA-SC code. This work is completed by introduction of a virtual two-phase region near the pseudo-critical line. A smooth transition of void fraction can be realized. In addition, several heat transfer correlations right underneath the critical point are introduced into this code to capture the heat transfer behavior during the trans-critical transient. Some experimental data from simple geometry, e.g. the single tube, small rod bundle, is used to validate and evaluate this new developed COBRA-SC code. The predicted results show a good agreement with the experimental data, demonstrating good feasibility of this code for SCWR condition. A code to code comparison between COBRA-SC and ATHLET for a blowdown transient of a small fuel assembly is also presented and discussed in this paper

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-15

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

Pulse-shape discrimination techniques for the COBRA double beta-decay experiment at LNGS

Science.gov (United States)

Zatschler, S.; COBRA Collaboration

2017-09-01

In modern elementary particle physics several questions arise from the fact that neutrino oscillation experiments have found neutrinos to be massive. Among them is the so far unknown nature of neutrinos: either they act as so-called Majorana particles, where one cannot distinguish between particle and antiparticle, or they are Dirac particles like all the other fermions in the Standard Model. The study of neutrinoless double beta-decay (0νββ-decay), where the lepton number conservation is violated by two units, could answer the question regarding the underlying nature of neutrinos and might also shed light on the mechanism responsible for the mass generation. So far there is no experimental evidence for the existence of 0νββ-decay, hence, existing experiments have to be improved and novel techniques should be explored. One of the next-generation experiments dedicated to the search for this ultra-rare decay is the COBRA experiment. This article gives an overview of techniques to identify and reject background based on pulse-shape discrimination.

Application of best estimate thermalhydraulic codes for the safety analysis of research reactors

International Nuclear Information System (INIS)

Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.

2006-01-01

system codes like RELAP5, COBRA and MARS to the research reactor needs has been confirmed from recent works. Definitely, system codes are mature for application to transient analysis in research reactors. However, code limitations have been found in predicting pressure drops as a function of mass flux at low values of mass flux when nucleate boiling occurs. The importance of the Whittle and Forgan experiments shall be mentioned and therefore furthermore, code validation must be demonstrated for the range of parameters of interest to research reactors. (author)

Experimental Analysis of Steady-State Maneuvering Effects on Transmission Vibration Patterns Recorded in an AH-1 Cobra Helicopter

Science.gov (United States)

Huff, Edward M.; Dzwonczyk, Mark; Norvig, Peter (Technical Monitor)

2000-01-01

Flight experiment was designed primarily to determine the extent to which steady-state maneuvers influence characteristic vibration patterns measured at the input pinion and output annulus gear locations of the main transmission. If results were to indicate that maneuvers systematically influence vibration patterns, more extensive studies would be planned to explore the response surface. It was also designed to collect baseline data for comparison with experimental data to be recorded at a later date from test stands at Glenn Research Center. Finally, because this was the first vibration flight study on the Cobra aircraft, considerable energy was invested in developing an in-flight recording apparatus, as well as exploring acceleration mounting methods, and generally learning about the overall vibratory characteristics of the aircraft itself.

Staged Z-pinch experiments on the Mega-Ampere current driver COBRA

Science.gov (United States)

Valenzuela, Julio; Banasek, Jacob; Byvank, Thomas; Conti, Fabio; Greenly, John; Hammer, David; Potter, William; Rocco, Sophia; Ross, Michael; Wessel, Frank; Narkis, Jeff; Rahman, Hafiz; Ruskov, Emil; Beg, Farhat

2017-10-01

Experiments were conducted on the Cornell's 1 MA, 100 ns current driver COBRA with the goal of better understanding the Staged Z-pinch physics and validating MHD codes. We used a gas injector composed of an annular (1.2 cm radius) high atomic number (e.g., Ar or Kr) gas-puff and an on-axis plasma gun that delivers the ionized hydrogen target. Liner implosion velocity and stability were studied using laser shadowgraphy and interferometry as well as XUV imaging. From the data, the signature of the MRT instability and zippering effect can be seen, but time integrated X-ray imaging show a stable target plasma. A key component of the experiment was the use of optical Thomson scattering (TS) diagnostics to characterize the liner and target plasmas. By fitting the experimental scattered spectra with synthetic data, electron and ion temperature as well as density can be obtained. Preliminary analysis shows significant scattered line broadening from the plasma on-axis ( 0.5 mm diameter) which can be explained by either a low temperature H plasma with Te =Ti =75eV, or by a hot plasma with Ti =3keV, Te =350eV if an Ar-H mixture is present with an Ar fraction higher than 10%. Funded by the Advanced Research Projects Agency - Energy, DE-AR0000569.

Multi-dimensional boron transport modeling in subchannel approach: Part I. Model selection, implementation and verification of COBRA-TF boron tracking model

Energy Technology Data Exchange (ETDEWEB)

Ozdemir, Ozkan Emre, E-mail: ozdemir@psu.edu [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Avramova, Maria N., E-mail: mna109@psu.edu [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Sato, Kenya, E-mail: kenya_sato@mhi.co.jp [Mitsubishi Heavy Industries (MHI), Kobe (Japan)

2014-10-15

Highlights: ► Implementation of multidimensional boron transport model in a subchannel approach. ► Studies on cross flow mechanism, heat transfer and lateral pressure drop effects. ► Verification of the implemented model via code-to-code comparison with CFD code. - Abstract: The risk of reflux condensation especially during a Small Break Loss Of Coolant Accident (SB-LOCA) and the complications of tracking the boron concentration experimentally inside the primary coolant system have stimulated and subsequently have been a focus of many computational studies on boron tracking simulations in nuclear reactors. This paper presents the development and implementation of a multidimensional boron transport model with Modified Godunov Scheme within a thermal-hydraulic code based on a subchannel approach. The cross flow mechanism in multiple-subchannel rod bundle geometry as well as the heat transfer and lateral pressure drop effects are considered in the performed studies on simulations of deboration and boration cases. The Pennsylvania State University (PSU) version of the COBRA-TF (CTF) code was chosen for the implementation of three different boron tracking models: First Order Accurate Upwind Difference Scheme, Second Order Accurate Godunov Scheme, and Modified Godunov Scheme. Based on the performed nodalization sensitivity studies, the Modified Godunov Scheme approach with a physical diffusion term was determined to provide the best solution in terms of precision and accuracy. As a part of the verification and validation activities, a code-to-code comparison was carried out with the STAR-CD computational fluid dynamics (CFD) code and presented here. The objective of this study was two-fold: (1) to verify the accuracy of the newly developed CTF boron tracking model against CFD calculations; and (2) to investigate its numerical advantages as compared to other thermal-hydraulics codes.

Temperature distribution determination of JPSR power reactor fuel element and cladding

International Nuclear Information System (INIS)

Sudarmono

1996-01-01

In order to utilize of fuel rod efficiency, a concept of JAERI passive Safety Reactor (JPSR) has been developed in Japan Atomic Energy Research Institute. In the JPSR design, UO 2 . are adopted as a fuel rod. The temperature distribution in the fuel rod and cladding in the hottest channel is a potential limiting design constraint of the JPSR. In the present determination, temperature distribution of the fuel rod and cladding for JPSR were PET:formed using COBRA-IV-I to evaluate the safety margin of the present JPSR design. In this method, the whole core was represented by the 1/4 sector and divided into 50 subchannels and 40 axial nodes. The temperature become maximum at the elevation of 1.922 and 2.196 m in the typical cell under operating condition. The maximum temperature in the center of the fuel rod surface of the fuel rod and cladding were 1620,4 o C, 722,8 o C, and 348,6 o C. The maximum results of temperature in the center of the fuel rod and cladding; were 2015,28 o C and 550 o C which were observed at 3.1 second in the typical cell

The Antinociceptive Effects of Iranian Cobra Snake Venom using Formalin Test

Directory of Open Access Journals (Sweden)

Zahra Hadi Chegeni

2015-06-01

Full Text Available Abstract Background: There have been numerous reports of snake venoms being employed as analgesics in attempts to relieve severe pain associated with cancer, immune dysfunction and viral infections. This study investigates the antinociceptive effects of iranian cobra snake venom (Naja naja oxiana in comparison with morphine and lidocain on laboratorial femal mice. Materials and Methods: This study has been done on 48 NMRI female mice of 18-20 g in weight. Antinociceptive activeity of snake venom was evaluated by formalin test. In this test, the animals were divided into 6 groups (each group consisting of 8 mice: Sham, positive Control (receiving morphine at dose of 5 mg/kg, and receiving lidocain at dose of 20 mg/kg, and experimental groups receiving venom at doses of 1, 3 and 4/5 µg/mice. In all groups, the formalin test was recorded for 60 min after administration of venom and drugs in mice. Data were analyzed using one-way ANOVA and Tukey test. Results: The results showed that the venom of Naja naja oxiana decreased nociception meaningfully in both acute and chronic phases. We also showed that this venom revealed even a better analgesic activity in comparison with morphine and lidocain. Conclusion: This study showed that the antinociceptive effect of the venom was mediated through central nervous system and peripheral mechanisms. Although details of the mechanism remain unclear, and further studies should be considered to demonstrate its therapeutic effects.

CTF Theory Manual

Energy Technology Data Exchange (ETDEWEB)

Avramova, Maria N. [Pennsylvania State Univ., University Park, PA (United States); Salko, Robert K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-05-25

Coolant-Boiling in Rod Arrays|Two Fluids (COBRA-TF) is a thermal/ hydraulic (T/H) simulation code designed for light water reactor (LWR) vessel analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of 9 conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) and has been improved, updated, and subsequently re-branded as CTF. As part of the improvement process, it was necessary to generate sufficient documentation for the open-source code which had lacked such material upon being adopted by RDFMG. This document serves mainly as a theory manual for CTF, detailing the many two-phase heat transfer, drag, and important accident scenario models contained in the code as well as the numerical solution process utilized. Coding of the models is also discussed, all with consideration for updates that have been made when transitioning from COBRA-TF to CTF. Further documentation outside of this manual is also available at RDFMG which focus on code input deck generation and source code global variable and module listings.

MECHANISMS CONTROLLING SURFACE WATER QUALITY IN THE COBRAS RIVER SUB-BASIN, NORTHEASTERN BRAZIL

Directory of Open Access Journals (Sweden)

ALEXANDRE DE OLIVEIRA LIMA

2017-01-01

Full Text Available Stream water quality is dependent on many factors, including the source and quantity of the streamflow and the types of geology and soil along the path of the stream. This study aims to evaluate the origin and the mechanisms controlling the input of ions that effect surface water quality in the sub-basin of the Rio das Cobras, Rio Grande do Norte state, Northeastern Brazil. Thirteen ponds were identified for study: three in the main river and ten in the tributaries between, thus covering the whole area and lithology of the sub-basin. The samples were collected at two different times (late dry and rainy periods in the hydrological years 2009 and 2010, equating to total of four collection times. We analyzed the spatial and seasonal behavior of water quality in the sub-basin, using Piper diagrams, and analyzed the source of the ions using Guibbs diagram and molar ratios. With respect to ions, we found that water predominate in 82% sodium and 76% bicarbonate water (cations and anions, respectively. The main salinity control mechanism was related to the interaction of the colloidal particles (minerals and organic sediment with the ions dissolved in water. Based on the analysis of nitrates and nitrites there was no evidence of contamination from anthropogenic sources.

Influence of hirudin and cobra venom factor on the release of 14C-serotonin and 51chromium from human platelets induced by thrombin, collagen, aggregate gammaglobulin and HLA antibody

International Nuclear Information System (INIS)

Hagemeyer, G.M.

1982-01-01

The present work investigates the influence of hirudin and cobra venom factor on thrombin, collagen, aggregate gammaglobulin and HLA-antibody-induced release of 14 C-serotonin and 51 chromium from human platelets. Besides the platelet-specific release reaction ( 14 C-serotonin) the extent of platelet lysis was determined by measurement of the loss of 51 chromium from the platelets. The results showed the thrombin, collagen and aggregate-gammaglobulin-induced platelet alteration to be a non-complement-dependent reaction of the platelets with release of 14 C-serotonin. Following long-term incubation small quantities of 51 chromium are also released. As this release of 51 chromium cannot be inhibited using cobra venom factor and does not occur in washed platelets either, it is most probably a non-complement-dependent reaction. The HLA-antibody-induced, specific platelet alteration is both complement-dependent and complement-independent. Differentiation is possible by inhibition of the complement-dependent lysis. On the other hand thrombin is of no relevance to the collagen, aggregate gammaglobulin, and HLA-antibody-induced platelet alteration as the interactions of these substances with platelets are not inhibited by hirudin. The above results are confirmed by investigation of the 51 chromium uptake capacity of washed platelets treated previously with thrombin, collagen and HLA antibody. (orig./MG) [de

Implementation of the optimization for the methodology of the neutronic calculation and thermo-hydraulic in IEA-R1 reactor

International Nuclear Information System (INIS)

Stefani, Giovanni Laranjo de; Conti, Thadeu das Neves; Fedorenko, Giuliana G.; Castro, Vinicius A.; Maio, Mireia F.; Santos, Thiago Augusto dos

2011-01-01

This work objective was to create a manager program that would automate the programs and computer codes in use for neutronic calculation and thermo-hydraulic in IEA-R1 reactor thus making the process for calculation of safety parameters and for configuration change up to 98% faster than that used in the reactor today. This process was tested in combination with the reactor operators and is being implemented by the quality department. The main codes and programs involved in the calculations of configuration change are Leopard, Hammier-Technion, Twodb, Citation and Cobra. Calculations of delayed neutron and criticality coefficients given in the process of safety parameters calculation are given by the Hammer-Technion and Citation in a process that involves about eleven repetitions so that it meets all the necessary conditions (such different temperatures of the moderator and fuel). The results are entirely consistent with the expected and absolutely the same as those given by manual process. Thus the work shows its reliability as well the advantage of saving time, once a process that could take up to four hours was turned in one that takes around five minutes when done in a home computer. Much of this advantage is due to the fact that were created subprograms to treat the output of each program used and transform them into the input of the other programs, removing from it the intermediate essential data for this to occur, thus avoiding also a possible human error by handling the various data supplied. (author)

The role of COBRA-LIKE 2 function, as part of the complex network of interacting pathways regulating Arabidopsis seed mucilage polysaccharide matrix organization.

Science.gov (United States)

Ben-Tov, Daniela; Idan-Molakandov, Anat; Hugger, Anat; Ben-Shlush, Ilan; Günl, Markus; Yang, Bo; Usadel, Björn; Harpaz-Saad, Smadar

2018-05-01

The production of hydrophilic mucilage along the course of seed coat epidermal cell differentiation is a common adaptation in angiosperms. Previous studies have identified COBRA-LIKE 2 (COBL2), a member of the COBRA-LIKE gene family, as a novel component required for crystalline cellulose deposition in seed coat epidermal cells. In recent years, Arabidopsis seed coat epidermal cells (SCEs), also called mucilage secretory cells, have emerged as a powerful model system for the study of plant cell wall components biosynthesis, secretion, assembly and de muro modification. Despite accumulating data, the molecular mechanism of COBL function remains largely unknown. In the current research, we utilized genetic interactions to study the role of COBL2 as part of the protein network required for seed mucilage production. Using correlative phenotyping of structural and biochemical characteristics, unique features of the cobl2 extruded mucilage are revealed, including: 'unraveled' ray morphology, loss of primary cell wall 'pyramidal' organization, reduced Ruthenium red staining intensity of the adherent mucilage layer, and increased levels of the monosaccharides arabinose and galactose. Examination of the cobl2cesa5 double mutant provides insight into the interface between COBL function and cellulose deposition. Additionally, genetic interactions between cobl2 and fei1fei2 as well as between each of these mutants to mucilage-modified 2 (mum2) suggest that COBL2 functions independently of the FEI-SOS pathway. Altogether, the presented data place COBL2 within the complex protein network required for cell wall deposition in the context of seed mucilage and introduce new methodology expending the seed mucilage phenotyping toolbox. © 2018 The Authors The Plant Journal © 2018 John Wiley & Sons Ltd.

Selection and benchmarking of computer codes for research reactor core conversions

Energy Technology Data Exchange (ETDEWEB)

Yilmaz, Emin [School of Aerospace, Mechanical and Nuclear Engineering, University of Oklahoma, Norman, OK (United States); Jones, Barclay G [Nuclear Engineering Program, University of IL at Urbana-Champaign, Urbana, IL (United States)

1983-09-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC{sup 2}, COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of k{sub eff} is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

Selection and benchmarking of computer codes for research reactor core conversions

International Nuclear Information System (INIS)

Yilmaz, Emin; Jones, Barclay G.

1983-01-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC 2 , COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of k eff is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

Discrimination of alpha particles in CdZnTe detectors with coplanar grid for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Rebber, Henning [Universitaet Hamburg, Institut fuer Experimentalphysik, Luruper Chaussee 149, 22761 Hamburg (Germany); Collaboration: COBRA-Collaboration

2016-07-01

The aim of the COBRA experiment is the search for neutrinoless double beta decay using CdZnTe semiconductor detectors. A background rate in the order of 10{sup -3} counts per keV, kg and year is intended in order to be sensitive to a half-life larger than 10{sup 26} years. Measurements from a demonstrator setup and Monte Carlo simulations indicate that a large background component is due to alpha particles. These generate charge clouds of only few μm in diameter in the detector, leading to characteristic pulse features. Parameter-based cut criteria were developed to discriminate alpha events by means of their pulse shapes. The cuts were tested on data from alpha and beta irradiation of a (1 x 1 x 1) cm{sup 3} CdZnTe detector with coplanar grid. The pulse shapes of all event signals were read out by FADCs with a sampling rate of 100 MHz. The signals were reproduced by a detector simulation which hence was used to study the cuts for energies up to 3 MeV and different detector regions.

Final report V1.0 for the CORE Organic II funded project: Coordinating Organic Breeding Activities for Diversity - COBRA

DEFF Research Database (Denmark)

Pearce, Bruce; Kir, Alev; Andersen, Rikke Thomle

variation in climate and weather. In this context, COBRA aimed to support and develop organic plant breeding and seed production with a focus on increasing the use and potential of plant material with high genetic diversity in cereals (wheat and barley) and grain legumes (pea and faba bean), through...... ensuring seed quality and health Progress was made in handling individual seeds in terms of their actual and potential resistance to seed-borne disease. One of the most important problems, bunt of wheat, was advanced considerably in terms of the 'gene for gene' interaction between host and pathogen...... and understanding the resilience of the performance of composite cross populations of wheat. A wide range of molecular markers were identified in barley which will help in selecting genotypes adapted to expected future changes in climate and weather. Progress was also made with organic trials of grain legumes...

Isolation, expression and characterization of a novel dual serine protease inhibitor, OH-TCI, from king cobra venom.

Science.gov (United States)

He, Ying-Ying; Liu, Shu-Bai; Lee, Wen-Hui; Qian, Jin-Qiao; Zhang, Yun

2008-10-01

Snake venom Kunitz/BPTI members are good tools for understanding of structure-functional relationship between serine proteases and their inhibitors. A novel dual Kunitz/BPTI serine proteinase inhibitor named OH-TCI (trypsin- and chymotrypsin-dual inhibitor from Ophiophagus hannah) was isolated from king cobra venom by three chromatographic steps of gel filtration, trypsin affinity and reverse phase HPLC. OH-TCI is composed of 58 amino acid residues with a molecular mass of 6339Da. Successful expression of OH-TCI was performed as the maltose-binding fusion protein in E. coli DH5alpha. Much different from Oh11-1, the purified native and recombinant OH-TCI both had strong inhibitory activities against trypsin and chymotrypsin although the sequence identity (74.1%) between them is very high. The inhibitor constants (K(i)) of recombinant OH-TCI were 3.91 x 10(-7) and 8.46 x10(-8)M for trypsin and chymotrypsin, respectively. To our knowledge, it was the first report of Kunitz/BPTI serine proteinase inhibitor from snake venom that had equivalent trypsin and chymotrypsin inhibitory activities.

The Crystal Structure of Cobra Venom Factor, a Cofactor for C3- and C5-Convertase CVFBb

Energy Technology Data Exchange (ETDEWEB)

Krishnan, Vengadesan; Ponnuraj, Karthe; Xu, Yuanyuan; Macon, Kevin; Volanakis, John E.; Narayana, Sthanam V.L.; (Madras); (UAB)

2009-05-26

Cobra venom factor (CVF) is a functional analog of human complement component C3b, the active fragment of C3. Similar to C3b, in human and mammalian serum, CVF binds factor B, which is then cleaved by factor D, giving rise to the CVFBb complex that targets the same scissile bond in C3 as the authentic complement convertases C4bC2a and C3bBb. Unlike the latter, CVFBb is a stable complex and an efficient C5 convertase. We solved the crystal structure of CVF, isolated from Naja naja kouthia venom, at 2.6 {angstrom} resolution. The CVF crystal structure, an intermediate between C3b and C3c, lacks the TED domain and has the CUB domain in an identical position to that seen in C3b. The similarly positioned CUB and slightly displaced C345c domains of CVF could play a vital role in the formation of C3 convertases by providing important primary binding sites for factor B.

[Sequencing and analysis of the complete mitochondrial genome of the King Cobra, Ophiophagus hannah (Serpents: Elapidae)].

Science.gov (United States)

Chen, Nian; Lai, Xiao-Ping

2010-07-01

We obtained the complete mitochondrial genome of King Cobra(GenBank accession number: EU_921899) by Ex Taq-PCR, TA-cloning and primer-walking methods. This genome is very similar to other vertebrate, which is 17 267 bp in length and encodes 38 genes (including 13 protein-coding, 2 ribosomal RNA and 23 transfer RNA genes) and two long non-coding regions. The duplication of tRNA-Ile gene forms a new mitochondrial gene rearrangement model. Eight tRNA genes and one protein genes were transcribed from L strand, and the other genes were transcribed genes from H strand. Genes on the H strand show a fairly similar content of Adenosine and Thymine respectively, whereas those on the L strand have higher proportion of A than T. Combined rDNA sequence data (12S+16S rRNA) were used to reconstruct the phylogeny of 21 snake species for which complete mitochondrial genome sequences were available in the public databases. This large data set and an appropriate range of outgroup taxa demonstrated that Elapidae is more closely related to colubridae than viperidae, which supports the traditional viewpoints.

In-flight measurements of propeller blade deformation on a VUT100 cobra aeroplane using a co-rotating camera system

Science.gov (United States)

Boden, F.; Stasicki, B.; Szypuła, M.; Ružička, P.; Tvrdik, Z.; Ludwikowski, K.

2016-07-01

Knowledge of propeller or rotor blade behaviour under real operating conditions is crucial for optimizing the performance of a propeller or rotor system. A team of researchers, technicians and engineers from Avia Propeller, DLR, EVEKTOR and HARDsoft developed a rotating stereo camera system dedicated to in-flight blade deformation measurements. The whole system, co-rotating with the propeller at its full speed and hence exposed to high centrifugal forces and strong vibration, had been successfully tested on an EVEKTOR VUT 100 COBRA aeroplane in Kunovice (CZ) within the project AIM2—advanced in-flight measurement techniques funded by the European Commission (contract no. 266107). This paper will describe the work, starting from drawing the first sketch of the system up to performing the successful flight test. Apart from a description of the measurement hardware and the applied IPCT method, the paper will give some impressions of the flight test activities and discuss the results obtained from the measurements.

COBRA encodes a putative GPI-anchored protein, which is polarly localized and necessary for oriented cell expansion in Arabidopsis.

Science.gov (United States)

Schindelman, G; Morikami, A; Jung, J; Baskin, T I; Carpita, N C; Derbyshire, P; McCann, M C; Benfey, P N

2001-05-01

To control organ shape, plant cells expand differentially. The organization of the cellulose microfibrils in the cell wall is a key determinant of differential expansion. Mutations in the COBRA (COB) gene of Arabidopsis, known to affect the orientation of cell expansion in the root, are reported here to reduce the amount of crystalline cellulose in cell walls in the root growth zone. The COB gene, identified by map-based cloning, contains a sequence motif found in proteins that are anchored to the extracellular surface of the plasma membrane through a glycosylphosphatidylinositol (GPI) linkage. In animal cells, this lipid linkage is known to confer polar localization to proteins. The COB protein was detected predominately on the longitudinal sides of root cells in the zone of rapid elongation. Moreover, COB RNA levels are dramatically upregulated in cells entering the zone of rapid elongation. Based on these results, models are proposed for the role of COB as a regulator of oriented cell expansion.

An Alternative Inactivant for Rift Valley Fever Virus using Cobra Venom-derived L-Amino Oxidase, which is Related to its Immune Potential

Directory of Open Access Journals (Sweden)

Ebtesam M Al-Olayan

Full Text Available ABSTRACT Vaccine improvement depends on the formulation, adjuvant type and inactivant used. The type of formulation may interfere with immunogenicity. The present work aimed to evaluate the inactivation activity and related immune potential of the Cobra venom-derived LAO enzyme compared to the currently used inactivants (BPL and formalin for both animal and human vaccines. The RVF virus was completely inactivated within 6 hrs, 4 hrs and 2 hrs after treatment with Formalin, LAO and BPL, respectively. The vaccine potency [ED50] was arranged in a descending order from formalin (0.016 to BPL (0.005 and LAO (0.002. The total IgG levels, Neutralizing Index (NI and Interferon levels were significantly increased compared to those detected after immunization with the BPL- and Formalin-inactivated vaccine candidates.

Effects of sleeve blockages on axial velocity and intensity of turbulence in an unheated 7 x 7 rod bundle

International Nuclear Information System (INIS)

Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

1976-01-01

An experimental study is described which was performed to investigate the turbulent flow phenomena near postulated sleeve blockages in a model nuclear fuel rod bundle. The sleeve blockages were characteristic of fuel clad ''swelling'' or ''ballooning'' which could occur during loss-of-coolant accidents (LOCA) in pressurized water reactors. The study was conducted to provide information relative to the flow phenomena near postulated blockages to support detailed safety analyses of LOCAs. The results of the study are especially useful for verification of the hydraulic treatment of reactor core computer programs such as COBRA

Numerically induced pressure excursions in two-phase-flow calculations. Final report

International Nuclear Information System (INIS)

Mahaffy, J.H.; Liles, D.R.

1983-01-01

Pressure spikes that cannot be traced to any physical origin sometimes are observed when standard Eulerian finite-difference methods are used to calculate two-phase-flow transients. This problem occurs with varying frequency in nuclear reactor safety codes such as RELAP, RETRAN, COBRA, and TRAC. These spikes usually result from numerical water packing or from interactions between spatial discretization and heat transfer

Growth performance, productivity and diseases susceptibility of barley varieties in Slovenia within the Cobra project’s site comparison

Directory of Open Access Journals (Sweden)

Grobelnik Mlakar Silva

2016-12-01

Full Text Available Different plant genotypes react differently in different climates. A field experiment was carried out to estimate the growth performance, productivity and diseases susceptibility of spring barley varieties in the Slovenian climate. We received some varieties, mainly of Nordic origin, from the Technical University of Denmark, a COBRA project partner, which were previously tested in estimated future climate in RERAF phytotron. Varieties of the highest grain yield (3,993 kg ha-1 in ‘Evergreen’ to 5,146 kg ha-1 in ‘Sebastian’ were rather shorter (58.7 cm to 67.1 cm and mostly had the highest specific grain weight (54.3 to 58.6 kg 100 L-1 and 1000-kernel weight (30.2 to 37.1 g. They developed 1,561 to 2,532 tillers m-2 and 515 to 840 ears m-2 and reached a heading stage between 13th and 25th of May. The tested varieties seem rather insusceptible to most common diseases, but susceptible to cereal leaf beetle attacks.

CASL VMA FY16 Milestone Report (L3:VMA.VUQ.P13.07) Westinghouse Mixing with COBRA-TF

Energy Technology Data Exchange (ETDEWEB)

Gordon, Natalie [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-09-30

COBRA-TF (CTF) is a low-resolution code currently maintained as CASL's subchannel analysis tool. CTF operates as a two-phase, compressible code over a mesh comprised of subchannels and axial discretized nodes. In part because CTF is a low-resolution code, simulation run time is not computationally expensive, only on the order of minutes. Hi-resolution codes such as STAR-CCM+ can be used to train lower-fidelity codes such as CTF. Unlike STAR-CCM+, CTF has no turbulence model, only a two-phase turbulent mixing coefficient, β. β can be set to a constant value or calculated in terms of Reynolds number using an empirical correlation. Results from STAR-CCM+ can be used to inform the appropriate value of β. Once β is calibrated, CTF runs can be an inexpensive alternative to costly STAR-CCM+ runs for scoping analyses. Based on the results of CTF runs, STAR-CCM+ can be run for specific parameters of interest. CASL areas of application are CIPS for single phase analysis and DNB-CTF for two-phase analysis.

Three-dimensional coupled kinetics/thermal- hydraulic benchmark TRIGA experiments

International Nuclear Information System (INIS)

Feltus, Madeline Anne; Miller, William Scott

2000-01-01

This research project provides separate effects tests in order to benchmark neutron kinetics models coupled with thermal-hydraulic (T/H) models used in best-estimate codes such as the Nuclear Regulatory Commission's (NRC) RELAP and TRAC code series and industrial codes such as RETRAN. Before this research project was initiated, no adequate experimental data existed for reactivity initiated transients that could be used to assess coupled three-dimensional (3D) kinetics and 3D T/H codes which have been, or are being developed around the world. Using various Test Reactor Isotope General Atomic (TRIGA) reactor core configurations at the Penn State Breazeale Reactor (PSBR), it is possible to determine the level of neutronics modeling required to describe kinetics and T/H feedback interactions. This research demonstrates that the small compact PSBR TRIGA core does not necessarily behave as a point kinetics reactor, but that this TRIGA can provide actual test results for 3D kinetics code benchmark efforts. This research focused on developing in-reactor tests that exhibited 3D neutronics effects coupled with 3D T/H feedback. A variety of pulses were used to evaluate the level of kinetics modeling needed for prompt temperature feedback in the fuel. Ramps and square waves were used to evaluate the detail of modeling needed for the delayed T/H feedback of the coolant. A stepped ramp was performed to evaluate and verify the derived thermal constants for the specific PSBR TRIGA core loading pattern. As part of the analytical benchmark research, the STAR 3D kinetics code (, STAR: Space and time analysis of reactors, Version 5, Level 3, Users Guide, Yankee Atomic Electric Company, YEAC 1758, Bolton, MA) was used to model the transient experiments. The STAR models were coupled with the one-dimensional (1D) WIGL and LRA and 3D COBRA (, COBRA IIIC: A digital computer program for steady-state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements, Battelle

Control rod drop transient analysis with the coupled parallel code pCTF-PARCSv2.7

International Nuclear Information System (INIS)

Ramos, Enrique; Roman, Jose E.; Abarca, Agustín; Miró, Rafael; Bermejo, Juan A.

2016-01-01

Highlights: • An MPI parallel version of the thermal–hydraulic subchannel code COBRA-TF has been developed. • The parallel code has been coupled to the 3D neutron diffusion code PARCSv2.7. • The new codes are validated with a control rod drop transient. - Abstract: In order to reduce the response time when simulating large reactors in detail, a parallel version of the thermal–hydraulic subchannel code COBRA-TF (CTF) has been developed using the standard Message Passing Interface (MPI). The parallelization is oriented to reactor cells, so it is best suited for models consisting of many cells. The generation of the Jacobian matrix is parallelized, in such a way that each processor is in charge of generating the data associated with a subset of cells. Also, the solution of the linear system of equations is done in parallel, using the PETSc toolkit. With the goal of creating a powerful tool to simulate the reactor core behavior during asymmetrical transients, the 3D neutron diffusion code PARCSv2.7 (PARCS) has been coupled with the parallel version of CTF (pCTF) using the Parallel Virtual Machine (PVM) technology. In order to validate the correctness of the parallel coupled code, a control rod drop transient has been simulated comparing the results with the real experimental measures acquired during an NPP real test.

Operation of CdZnTe Semiconductor Detectors in Liquid Scintillator for the COBRA Experiment

International Nuclear Information System (INIS)

Oldorf, Christian

2015-08-01

COBRA, the Cadmium-Zinc-Telluride O-neutrino double-Beta Research Apparatus, is an experiment aiming for the measurement of the neutrinoless double beta decay with several isotopes, in particular 116 Cd, 106 Cd and 130 Te. A highly granular large scale experiment with about 400 kg of CdZnTe semiconductor detectors is currently under development. To provide evidence for the neutrinoless double beta decay of 116 Cd, a background rate in the order of 10 -3 counts/keV/kg/a is needed to achieve the required half-life sensitivity of at least 2 . 10 26 years. To reach this target, the detectors have to be operated in a highly pure environment, shielded from external radiation. Liquid scintillator is a promising candidate as a circum fluent replacement for the currently used lacquer. Next to the function as highly pure passivation material, liquid scintillator also acts as a neutron shield and active veto for external gammas. Within this thesis, the design, construction and assembly of a test set-up is described. The operation of four CdZnTe detectors after several years of storage in liquid scintillator is demonstrated. Next to extensive material compatibility tests prior to the assembly, the commissioning of the set-up and the characterization of the detectors are shown. Finally, results concerning the background reduction capability of liquid scintillator and the detection of cosmic muons are presented and compared to a Monte Carlo simulation.

Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

International Nuclear Information System (INIS)

Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

1982-01-01

The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

In vivo studies on detoxifying actions of aqueous bark extract of Prosopis cineraria against crude venom from Indian cobra (Naja naja

Directory of Open Access Journals (Sweden)

Thirunavukkarasu Sivaraman

2013-12-01

Full Text Available Detoxification effect of aqueous, methanol and petroleum ether extracts of medicinal plants such as Aristolochia bracteolata, Mucuna pruriens, Prosopis cineraria and Rauvolfia tetraphylla was systematically screened against lethality of crude venom of Naja naja using Swiss albino mice as animal models. We have herein demonstrated that aqueous bark extract of P. cineraria has substantial anti-venom potential vis-à-vis other extracts used in the present study. The aqueous extract at the dose of 14 mg/kg b.w. was able to almost completely neutralize the lethal activity of 3LD50 (1.12 mg/kg b.w. of the cobra venom and the extract did not cause any types of adverse side-effects to the animal models. The investigation justifies not only the veraciousness of the extract used by traditional healers of Asian subcontinent as antidotes to snake venoms and also suggests that the aqueous extract should contain specific inhibitors to most principle toxic components of the crude venom.

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

On the frontier of boiling curve and beyond design of its origin

International Nuclear Information System (INIS)

Stosic, Z.V.

2005-01-01

An advanced approach of Extended Design of the Boiling Curve beyond its origin is proposed. It is developed from the fact that both CHF (Critical Heat Flux) and rewetting affect the Boiling Curve on the heating surface through two simultaneous processes taking place on both sides of the heating surface. The first is two-phase flow thermal-hydraulics with resultant heat transferred from the heating surface to the coolant. The second one is the heat conduction through material itself, allied with the balance of generated and accumulated energy. Both of these processes are triggered by the change in HTC (Heat Transfer Coefficient) on the heating surface, which accordingly influences the Boiling Curve. Depending on direction of the Transition - from nucleate to film boiling or vice versa - these processes act differently and direct the Boiling Curve to diverse paths. The proposed physically based concept recognises this fact and introduces HTC as the triggering parameter with instant effect. It is implemented in the subchannel code COBRA 3-CP providing stable rewetting which has been deficient in COBRA since its origin. Results of validation and obtained agreements with transient measured data prove legality of the advanced concept of Boiling Curve. This approach is being used for transient analyses of PWR (Pressurised Water Reactor) gaining benefits from properly predicting the rewetting. The method is well-qualified to be applied also in other thermal-hydraulic codes like COBRA/TRAC, COBRA-TF, TRAC and/or RELAP, where the classical steady-state and poolboiling approach has been originally implemented. (author)

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Atmospheric Oxidation of Squalene: Molecular Study Using COBRA Modeling and High-Resolution Mass Spectrometry

Energy Technology Data Exchange (ETDEWEB)

Fooshee, David R.; Aiona, Paige K.; Laskin, Alexander; Laskin, Julia; Nizkorodov, Sergey; Baldi, Pierre

2015-10-22

Squalene is a major component of skin and plant surface lipids, and is known to be present at high concentrations in indoor dust. Its high reactivity toward ozone makes it an important ozone sink and a natural protectant against atmospheric oxidizing agents. While the volatile products of squalene ozonolysis are known, the condensed-phase products have not been characterized. We present an analysis of condensed-phase products resulting from an extensive oxidation of squalene by ozone probed by electrospray ionization (ESI) high-resolution mass spectrometry (HR-MS). A complex distribution of nearly 1,300 peaks assignable to molecular formulas is observed in direct infusion positive ion mode ESI mass spectra. The distribution of peaks in the mass spectra suggests that there are extensive cross-coupling reactions between hydroxy-carbonyl products of squalene ozonolysis. To get additional insights into the mechanism, we apply a Computational Brewing Application (COBRA) to simulate the oxidation of squalene in the presence of ozone, and compare predicted results with those observed by the HR-MS experiments. The system predicts over one billion molecular structures between 0-1450 Da, which correspond to about 27,000 distinct elemental formulas. Over 83% of the squalene oxidation products inferred from the mass spectrometry data are matched by the simulation. Simulation indicates a prevalence of peroxy groups, with hydroxyl and ether groups being the second-most important O-containing functional groups formed during squalene oxidation. These highly oxidized products of squalene ozonolysis may accumulate on indoor dust and surfaces, and contribute to their redox capacity.

Development of an artificial neural network model for on-line thermal margin estimation of a nuclear reactor core

International Nuclear Information System (INIS)

Kim, Hyun Koon

1992-02-01

One of the key safety parameters related to thermal margin in a Pressurized Water Reactor (PWR) core, is Departure from Nucleate Boiling Ratio (DNBR), which is to be assessed and continuously monitored during operation via either an analog or a digital monitoring system. The digital monitoring system, in general, allows more thermal margin than the analog system through the on-line computation of DNBR using the measured parameters as inputs to a simplified, fast running computer code. The purpose of this thesis is to develop an advanced method for on-line DNBR estimation by introducing an artifactual neural network model for best-estimation of DNBR at the given reactor operating conditions. the neural network model, consisting of three layers with five operating parameters in the input layer, provides real-time prediction accuracy of DNBR by training the network against the detailed simulation results for various operating conditions. The overall training procedure is developed to learn the characteristics of DNBR behaviour in the reactor core. First, a set of random combination of input variables is generated by Latin Hypercube Sampling technique performed on a wide range of input parameters. Second, the target values of DNBR to be referenced for training are calculated using a detailed simulation code, COBRA-IV. Third, the optimized training input data are selected. Then, training is performed using an Error Back Propagation algorithm. After completion of training, the network is tested on the examining data set in order to investigate the generalization capability of the network responses for the steady state operating condition as well as for the transient situations where DNB is of a primary concern. The test results show that the values of DNBR predicted by the neural network are maintained at a high level of accuracy for the steady state condition, and are in good agreements with the transient situation, although slightly conservative as compared to those

Characterization of large volume CdZnTe detectors with a quad-grid structure for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Rohatsch, Katja [TU Dresden, Institut fuer Kern- und Teilchenphysik, 01069 Dresden (Germany); Collaboration: COBRA-Collaboration

2016-07-01

The COBRA experiment uses room temperature semiconductor detectors made of Cadmium-Zinc-Telluride, which contains several double beta isotopes, to search for neutrinoless double beta-decay. To compensate for poor hole transport in CdZnTe the detectors are equipped with a coplanar grid (CPG) instead of a planar anode. Currently, a demonstrator setup consisting of 64 1 cm{sup 3} CPG-detectors is in operation at the LNGS in Italy to prove the concept and to determine the long-term stability of the detectors and the instrumentation. For a future large scale experiment it is planned to use larger CdZnTe detectors with a volume of 6 cm{sup 3}, because of the better surface-to-volume ratio and the higher full energy detection efficiency. This will also reduce the background contribution of surface contaminations. Before the installation at the LNGS the new detector design is validated and studied in detail. This talk presents a laboratory experiment for the characterization with γ-radiation of 6 cm{sup 3} CdZnTe quad-grid detectors. The anode of such a detector is divided into four sub-CPGs. The characterization routine consists of the determination of the optimal working point and two-dimensional spatially resolved scans with a highly collimated γ-source.

Intoxicação por veneno de cobra: necrose symetrica da cortex renal: uremia

Directory of Open Access Journals (Sweden)

A. Penna de Azevedo

1938-01-01

Full Text Available Em um caso fatal de ophidismo, em individuo de 15 annos de edade, picado por uma cobra jararaca (Bothrops jararaca na face externa da perna direita e que veio a fallecer 26 dias apoz o accidente, os A.A, descrevem as lesões anatomo-pathologicas encontradas e as modificações do metabolismo, evidenciadas pelos exames chimicos do sangue. As principaes alterações existentes, acham-se localisadas nos rins os quaes apresentam lesões de glomerulonephrite diffusa e o aspecto typico da necrose cortical symmetrica. Como alterações de maior significação observam-se ainda lesões vasculares de grande intensidade e constituidas essencialmente por processo de endoarterite productiva. A necrose symmetrica da cortex renal, a vista das intensas alterações vasculares (endoarterite productiva que acarretaram a obliteração das arterías, é considerada como a consequencia immediata de taes lesões vasculares. Os vasos renaes, séde do processo inflammatorio, são as arterias interlobar, arciforme e interlobular, mas principalmente as arteriolares da camada cortical. O processo de endoarterite assume sempre o carater progressivo, de modo que a luz vascular vae sendo aos poucos, totalmente obstruida. Ao contrario do que se tem observado nos casos de necrose cortical symmetrica, citados na literatura, em que as alterações parenchymatosas são consequentes a thrombose dos vasos reanes, no caso presente esse aspecto não foi verificado mas tão sómente a existencia da endoarterite productiva obliterante. Consideram os A.A. as lesões renaes no caso que estudaram, como a resultante da actuação lenta e prolongada do veneno de cobra sobre as estructuras renaes, baseados nos seguintes factos já conhecidos e admittidos: eliminação do veneno de cobra pelos rins; capacidade do mesmo veneno, determinar a glomerulo-nephrite diffusa e acção do veneno de cobra sobre o endothelio vascular, facilitada essencialmente pela funcção especifica do orgão. As

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Avaliação de características dos regimes de umidade na flona de Caxiuanã-PA durante o experimento COBRA-PARÁ Evaluation of characteristics of the moisture regimes in Caxiuanã national forest during COBRA-PARÁ experiment

Directory of Open Access Journals (Sweden)

Ludmila Monteiro da Silva

2010-03-01

Full Text Available Procura-se investigar a validade de um método de classificação de regimes de umidade, baseado na caracterização de diferentes "estados" da Camada Limite Atmosférica Tropical (CLAT, acima de uma área de floresta, de acordo com a metodologia proposta por Mahrt (1991. Para essas análises foram utilizados dados de radiossondagens e de uma torre micrometeorológica, coletados durante o período menos chuvoso da região, obtidos durante o experimento "COBRA-PARÁ" (realizado no período de 30/10 a 15/11 de 2006. A análise dos regimes de umidade consiste na representação em espaço de fase dos dados disponíveis da razão de Bowen (β, em função do parâmetro -h/L (onde h é a altura da camada de mistura turbulenta e L é o comprimento de Obukhov. Dependendo da localização dos dados nesse espaço foi possível caracterizar as seguintes classes: classe I - ar seco e instável; classe II - vento seco predominante; classe III - vento úmido; classe IV - condição úmida e instável; classe V - condensação de vapor d'água na superfície; classe VI - condição estável dominante; e classe VII -formação de orvalho induzido por radiação noturna resfriando a superfície. Das classes mencionadas, aquelas mais freqüentemente observadas em Caxiuanã, foram as III, IV e VI.We investigate the validity of a method of humidity regimes classification, based on different "states" characterization of the Tropical Atmospheric Boundary Layer (TABL, above a forest area, according to the methodology proposed by Mahrt (1991. To perform this investigation we used radiosonde information and micrometeorological tower data collected during the drier season of the region, during the experiment "COBRA-PARÁ" (carried out from 30/10 to 15/11, 2006. The analysis of moisture regimes is based on the "phase space" data representation, where the Bowen ratio (β is plotted against the -h/L parameter (where h is the height of the turbulent mixing layer and L is

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

COBRA: A prospective multimodal imaging study of dopamine, brain structure and function, and cognition.

Science.gov (United States)

Nevalainen, N; Riklund, K; Andersson, M; Axelsson, J; Ögren, M; Lövdén, M; Lindenberger, U; Bäckman, L; Nyberg, L

2015-07-01

Cognitive decline is a characteristic feature of normal human aging. Previous work has demonstrated marked interindividual variability in onset and rate of decline. Such variability has been linked to factors such as maintenance of functional and structural brain integrity, genetics, and lifestyle. Still, few, if any, studies have combined a longitudinal design with repeated multimodal imaging and a comprehensive assessment of cognition as well as genetic and lifestyle factors. The present paper introduces the Cognition, Brain, and Aging (COBRA) study, in which cognitive performance and brain structure and function are measured in a cohort of 181 older adults aged 64 to 68 years at baseline. Participants will be followed longitudinally over a 10-year period, resulting in a total of three equally spaced measurement occasions. The measurement protocol at each occasion comprises a comprehensive set of behavioral and imaging measures. Cognitive performance is evaluated via computerized testing of working memory, episodic memory, perceptual speed, motor speed, implicit sequence learning, and vocabulary. Brain imaging is performed using positron emission tomography with [(11)C]-raclopride to assess dopamine D2/D3 receptor availability. Structural magnetic resonance imaging (MRI) is used for assessment of white and gray-matter integrity and cerebrovascular perfusion, and functional MRI maps brain activation during rest and active task conditions. Lifestyle descriptives are collected, and blood samples are obtained and stored for future evaluation. Here, we present selected results from the baseline assessment along with a discussion of sample characteristics and methodological considerations that determined the design of the study. This article is part of a Special Issue entitled SI: Memory & Aging. Copyright © 2014 The Authors. Published by Elsevier B.V. All rights reserved.

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

Simulation of hot-channel transients for PHWR reactors

International Nuclear Information System (INIS)

Masriera, N.A.

1988-01-01

For the simulation of transients a whole-plant code is needed. These codes model the core in a very simplified way. When local variables have to be calculated a different kind of code is needed: a subchannel-code. This report studies the use of the cobra code as a subchannel-code, for the simulation of a PHWR fuel channel, considering that this code was developed for PWR cores calculation. A special effort is made to obtain optimized models for different calculations: steady state, soft transients and severe transients. These models differ in number of subchannels, axial nodes, and the choice of the most important variables. (Author) [es

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Development and assessment of multi-dimensional flow model in MARS compared with the RPI air-water experiment

International Nuclear Information System (INIS)

Lee, Seok Min; Lee, Un Chul; Bae, Sung Won; Chung, Bub Dong

2004-01-01

The Multi-Dimensional flow models in system code have been developed during the past many years. RELAP5-3D, CATHARE and TRACE has its specific multi-dimensional flow models and successfully applied it to the system safety analysis. In KAERI, also, MARS(Multi-dimensional Analysis of Reactor Safety) code was developed by integrating RELAP5/MOD3 code and COBRA-TF code. Even though COBRA-TF module can analyze three-dimensional flow models, it has a limitation to apply 3D shear stress dominant phenomena or cylindrical geometry. Therefore, Multi-dimensional analysis models are newly developed by implementing three-dimensional momentum flux and diffusion terms. The multi-dimensional model has been assessed compared with multi-dimensional conceptual problems and CFD code results. Although the assessment results were reasonable, the multi-dimensional model has not been validated to two-phase flow using experimental data. In this paper, the multi-dimensional air-water two-phase flow experiment was simulated and analyzed

RETRAN safety analyses of the nuclear-powered ship Mutsu

International Nuclear Information System (INIS)

Yoshinori, N.; Ishida, T.; Tanaka, Y.; Yoshiaki, F.

1983-01-01

A number of operational transient analyses of the nuclear-powered ship Mutsu have been performed in response to Japanese nuclear safety regulatory concerns. The RETRAN and COBRA-IV computer codes were used to provide a quantitative basis for the safety evaluation of the plant. This evaluation includes a complete loss of load without reactor scram, an excessive load increase incident, and an accidental depressurization of the primary system. The minimum departure from nucleate boiling ratio remained in excess of 1.53 for these three transients. Hence, the integrity of the core was shown to be maintained during these transients. Comparing the transient behaviors with those of land-based pressurized water reactors, the characteristic features of the Mutsu reactor were presented and the safety of the plant under the operational transient conditions was confirmed

Utilization of the MAT method to analyze the nucleate boiling boundary in rod bundles subchannels

International Nuclear Information System (INIS)

Pedron, M.Q.

1983-01-01

The digital program PANTERA-1P, a new version of the COBRA-IIIC code, developed at CDTN, is directed to the thermal-hydraulic analysis of water cooled rod bundles and reactor cores, insteady state and transient conditions. Both the new and the old code versions have identical capacities in what concerns evaluation of fluid variables, nevertheless PANTERA-1P has better and faster performance. Improvements introduced in the scheme for solution of the conservation equations have contributed significantly to reduce the computer time, without affecting the accuracy of results. While the momentum equations are solved in COBRA-IIIC for the crossflow distribution, the PANTERA-1P code solves these equations for the pressure distribution by using the MAT method (Modified and Advanced Theta). The calculation of the pressure coefficient matrix has been optimized and simultaneous linear equations are solved optionally by means of the transpose elimination with storage requirements or the successive over-relaxation methods. The program presents others features specially in what concerns the thermal conduction model for fuel rods and the critical heat flux calculations options. A new input/output scheme is provided for optional use of the British or Internacional System of Units. The results of the program are compared to the critical heat flux experimental data and to the results of COBRA-IIIC. Excellent agreement is observed in both cases. (Author) [pt

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

Effects of plant lectin from cobra lily, Arisaema curvatum Kunth on development of melon fruit fly, Bactrocera cucurbitae (Coq.).

Science.gov (United States)

Singh, Kuljinder; Kaur, Manpreet; Rup, Pushpinder J; Singh, Jatinder

2008-11-01

The lectin from tubers of cobra lily, Arisaema curvatum Kunth was purified by affinity chromatography using asialofetuin-linked amino activated porous silica beads. The concentration dependent effect of lectin was studied on second instar larvae (64-72 hr) of Bactrocera cucurbitae (Coq.). The treatment not only resulted in a significant reduction in the percentage pupation and emergence of the adults from treated larvae but it also prolonged the remaining larval development period. A very low LC50 value, 39 mgl(-1) of lectin was obtained on the basis of adult emergence using probit analysis. The activity of three hydrolase enzymes (esterases, acid and alkaline phosphatases), one oxidoreductase (catalase) and one group transfer enzyme (GSTs: Glutathione S-transferases) was assayed in second instar larvae under the influence of the LC50 of lectin at increasing exposure intervals (0, 24, 48 and 72 hr). The Arisaema curvatum lectin significantly decreased the activity of all the enzymes except for esterases, where the activity increased as compared to control at all exposure intervals. The decrease in pupation and emergence as well as significant suppression in the activities of two hydrolases, one oxidoreductase and one GST enzyme in treated larvae of B. cucurbitae indicated that this lectin has anti-metabolic effect on the melon fruit fly larvae.

Comprehensive Analysis of the COBRA-Like (COBL) Gene Family in Gossypium Identifies Two COBLs Potentially Associated with Fiber Quality

Science.gov (United States)

Niu, Erli; Shang, Xiaoguang; Cheng, Chaoze; Bao, Jianghao; Zeng, Yanda; Cai, Caiping; Du, Xiongming; Guo, Wangzhen

2015-01-01

COBRA-Like (COBL) genes, which encode a plant-specific glycosylphosphatidylinositol (GPI) anchored protein, have been proven to be key regulators in the orientation of cell expansion and cellulose crystallinity status. Genome-wide analysis has been performed in A. thaliana, O. sativa, Z. mays and S. lycopersicum, but little in Gossypium. Here we identified 19, 18 and 33 candidate COBL genes from three sequenced cotton species, diploid cotton G. raimondii, G. arboreum and tetraploid cotton G. hirsutum acc. TM-1, respectively. These COBL members were anchored onto 10 chromosomes in G. raimondii and could be divided into two subgroups. Expression patterns of COBL genes showed highly developmental and spatial regulation in G. hirsutum acc. TM-1. Of them, GhCOBL9 and GhCOBL13 were preferentially expressed at the secondary cell wall stage of fiber development and had significantly co-upregulated expression with cellulose synthase genes GhCESA4, GhCESA7 and GhCESA8. Besides, GhCOBL9 Dt and GhCOBL13 Dt were co-localized with previously reported cotton fiber quality quantitative trait loci (QTLs) and the favorable allele types of GhCOBL9 Dt had significantly positive correlations with fiber quality traits, indicating that these two genes might play an important role in fiber development. PMID:26710066

Preliminary analysis on incore performance of nuclear fuel: pt. 4

International Nuclear Information System (INIS)

Noh, S.K.; Chang, M.H.; Lee, C.C.; Chung, Y.H.; Kuk, K.Y.; Park, C.Y.; Lee, S.K.

1981-01-01

An analysis has been performed for thermal hydraulic design parameters of Wolsung-1 reactor core in steady state with the help of a computer code COBRA-IV-I. The design parameters are coolant enthalpy, flow velocity, coolant quality, pressure and fuel temperature distribution. The maximum power channel has been taken into account in this work. The results appear to be reasonably agreeable with data from PSR'S, with the maximum difference between this work and PSR'S being 4.3%

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Protein Characterization of Javan Cobra (Naja sputatrix) Venom Following Sun Exposure and Photo-Oxidation Treatment

Science.gov (United States)

Sulistiyani; Biki, R. S.; Andrianto, D.

2017-03-01

Snake venom has always been known for its toxicity that can cause fatality, however, it is also one of the important biological resources to be used for disease treatment. In Indonesia, snake venom previously expose under the sun has been used for alternative treatment of some diseases such as dengue fever, atherosclerosis, cancer, and diabetes. There has been very little scientific evidence on the use of snake venom of Indonesia origin as well as its protein characteristic. Thus, the objective of this research is to characterize the protein content and the specific activity of the venom of Javan Cobra (N.sputatrix) when treated with sun exposure in comparison with photo-oxidation by ultraviolet. Qualitative analysis of protein contents was determined using sodium dodecyl sulfate polyacrylamide gel electrophoresis (SDS PAGE). The L-amino acid oxidase activity (LAAO) and the phospholipase A2 (PLA2) activities were determined using spectrophotometry. The venom’s protein was separated into 5 main protein bands with molecular weight ranging from 14 to 108 kDa. A time course study showed that the venom lost 91% of its LAAO activity and 96% of PLA2 activity after 6 hours of sun exposure. UV photo-oxidation carried out for 3 hours decreased 91% of LAAO activity, and almost diminished all of PLA2 activity (99.8%). These findings suggest that the exposure of N. sputatrix venom under the sun and UV photo-oxidation decreased its toxicity as shown by the significant reduction of the enzymes activity, but did not affect the protein’s integrity. Therefore, these approaches produced N.sputatrix venom with less toxicity but still withheld other characters of intact proteins.

Brittle Culm1, a COBRA-Like Protein, Functions in Cellulose Assembly through Binding Cellulose Microfibrils

Science.gov (United States)

Zhang, Baocai; Liu, Xiangling; Yan, Meixian; Zhang, Lanjun; Shi, Yanyun; Zhang, Mu; Qian, Qian; Li, Jiayang; Zhou, Yihua

2013-01-01

Cellulose represents the most abundant biopolymer in nature and has great economic importance. Cellulose chains pack laterally into crystalline forms, stacking into a complicated crystallographic structure. However, the mechanism of cellulose crystallization is poorly understood. Here, via functional characterization, we report that Brittle Culm1 (BC1), a COBRA-like protein in rice, modifies cellulose crystallinity. BC1 was demonstrated to be a glycosylphosphatidylinositol (GPI) anchored protein and can be released into cell walls by removal of the GPI anchor. BC1 possesses a carbohydrate-binding module (CBM) at its N-terminus. In vitro binding assays showed that this CBM interacts specifically with crystalline cellulose, and several aromatic residues in this domain are essential for binding. It was further demonstrated that cell wall-localized BC1 via the CBM and GPI anchor is one functional form of BC1. X-ray diffraction (XRD) assays revealed that mutations in BC1 and knockdown of BC1 expression decrease the crystallite width of cellulose; overexpression of BC1 and the CBM-mutated BC1s caused varied crystallinity with results that were consistent with the in vitro binding assay. Moreover, interaction between the CBM and cellulose microfibrils was largely repressed when the cell wall residues were pre-stained with two cellulose dyes. Treating wild-type and bc1 seedlings with the dyes resulted in insensitive root growth responses in bc1 plants. Combined with the evidence that BC1 and three secondary wall cellulose synthases (CESAs) function in different steps of cellulose production as revealed by genetic analysis, we conclude that BC1 modulates cellulose assembly by interacting with cellulose and affecting microfibril crystallinity. PMID:23990797

Brittle Culm1, a COBRA-like protein, functions in cellulose assembly through binding cellulose microfibrils.

Directory of Open Access Journals (Sweden)

Lifeng Liu

Full Text Available Cellulose represents the most abundant biopolymer in nature and has great economic importance. Cellulose chains pack laterally into crystalline forms, stacking into a complicated crystallographic structure. However, the mechanism of cellulose crystallization is poorly understood. Here, via functional characterization, we report that Brittle Culm1 (BC1, a COBRA-like protein in rice, modifies cellulose crystallinity. BC1 was demonstrated to be a glycosylphosphatidylinositol (GPI anchored protein and can be released into cell walls by removal of the GPI anchor. BC1 possesses a carbohydrate-binding module (CBM at its N-terminus. In vitro binding assays showed that this CBM interacts specifically with crystalline cellulose, and several aromatic residues in this domain are essential for binding. It was further demonstrated that cell wall-localized BC1 via the CBM and GPI anchor is one functional form of BC1. X-ray diffraction (XRD assays revealed that mutations in BC1 and knockdown of BC1 expression decrease the crystallite width of cellulose; overexpression of BC1 and the CBM-mutated BC1s caused varied crystallinity with results that were consistent with the in vitro binding assay. Moreover, interaction between the CBM and cellulose microfibrils was largely repressed when the cell wall residues were pre-stained with two cellulose dyes. Treating wild-type and bc1 seedlings with the dyes resulted in insensitive root growth responses in bc1 plants. Combined with the evidence that BC1 and three secondary wall cellulose synthases (CESAs function in different steps of cellulose production as revealed by genetic analysis, we conclude that BC1 modulates cellulose assembly by interacting with cellulose and affecting microfibril crystallinity.

Brittle Culm1, a COBRA-like protein, functions in cellulose assembly through binding cellulose microfibrils.

Science.gov (United States)

Liu, Lifeng; Shang-Guan, Keke; Zhang, Baocai; Liu, Xiangling; Yan, Meixian; Zhang, Lanjun; Shi, Yanyun; Zhang, Mu; Qian, Qian; Li, Jiayang; Zhou, Yihua

2013-01-01

Cellulose represents the most abundant biopolymer in nature and has great economic importance. Cellulose chains pack laterally into crystalline forms, stacking into a complicated crystallographic structure. However, the mechanism of cellulose crystallization is poorly understood. Here, via functional characterization, we report that Brittle Culm1 (BC1), a COBRA-like protein in rice, modifies cellulose crystallinity. BC1 was demonstrated to be a glycosylphosphatidylinositol (GPI) anchored protein and can be released into cell walls by removal of the GPI anchor. BC1 possesses a carbohydrate-binding module (CBM) at its N-terminus. In vitro binding assays showed that this CBM interacts specifically with crystalline cellulose, and several aromatic residues in this domain are essential for binding. It was further demonstrated that cell wall-localized BC1 via the CBM and GPI anchor is one functional form of BC1. X-ray diffraction (XRD) assays revealed that mutations in BC1 and knockdown of BC1 expression decrease the crystallite width of cellulose; overexpression of BC1 and the CBM-mutated BC1s caused varied crystallinity with results that were consistent with the in vitro binding assay. Moreover, interaction between the CBM and cellulose microfibrils was largely repressed when the cell wall residues were pre-stained with two cellulose dyes. Treating wild-type and bc1 seedlings with the dyes resulted in insensitive root growth responses in bc1 plants. Combined with the evidence that BC1 and three secondary wall cellulose synthases (CESAs) function in different steps of cellulose production as revealed by genetic analysis, we conclude that BC1 modulates cellulose assembly by interacting with cellulose and affecting microfibril crystallinity.

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Opening of brain blood barrier induced by red light and central analgesic improvement of cobra neurotoxin.

Science.gov (United States)

Ye, Yong; Li, Yue; Fang, Fei

2014-05-05

Cobra neurotoxin (NT) has central analgesic effects, but it is difficult to pass through brain blood barrier (BBB). A novel method of red light induction is designed to help NT across BBB, which is based on photosensitizer activation by red light to generate reactive oxygen species (ROS) to open BBB. The effects were evaluated on cell models and animals in vivo with illumination by semiconductor laser at 670nm on photosensitizer pheophorbide isolated from silkworm excrement. Brain microvascular endothelial cells and astrocytes were co-cultured to build up BBB cell model. The radioactivity of (125)I-NT was measured in cells and tissues for NT permeation. Three ways of cranial irradiation, nasal cavity and intravascular irradiation were tested with combined injection of (125)I-NT 20μg/kg and pheophorbide 100μg/kg to rats, and organs of rats were separated and determined the radioactivity. Paw pressure test in rats, hot plate and writhing test in mice were applied to appraise the analgesic effects. NT across BBB cell model increased with time of illumination, and reached stable level after 60min. So did ROS in cells. NT mainly distributed in liver and kidney of rats, significantly increased in brain after illumination, and improved analgesic effects. Excitation of pheophorbide at red light produces ROS to open BBB, help NT enter brain, and enhance its central action. This research provides a new method for drug across BBB to improve its central role. Copyright © 2014 Elsevier B.V. All rights reserved.

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Avaliação da influência de tensoativos na pele de muda de cobra (Bothrops jararaca e Spilotis pullatus) por espectroscopia fatoacústica no infravermelho, calorimetria exploratória diferencial e espectroscopia Raman

OpenAIRE

Aurea Cristina Lemos Lacerda

2004-01-01

A influência dos tensoativos lauril sulfato de sódio, cloreto de cetil trimetil amônio e álcool láurico etoxilado com 12 moles de óxido de etileno sobre o stratum corneum da pele de muda das cobras Bothrops jararaca e Spilotis pullatus foi avaliada através das técnicas biofísicas de PAS-FTIR, FT-Raman e DSC. Foram utilizadas soluções dos tensoativos em concentrações acima e abaixo da cmc e tratamentos por 4 e 8 horas (stratum corneum íntegro) e por 12 horas (stratum corneum após a remoção mec...

Postirradiation examination of beryllium pebbles

International Nuclear Information System (INIS)

Gelles, D.S.

1998-01-01

Postirradiation examinations of COBRA-1A beryllium pebbles irradiated in the EBR-II fast reactor at neutron fluences which generated 2700--3700 appm helium have been performed. Measurements included density change, optical microscopy, scanning electron microscopy, and transmission electron microscopy. The major change in microstructure is development of unusually shaped helium bubbles forming as highly non-equiaxed thin platelet-like cavities on the basal plane. Measurement of the swelling due to cavity formation was in good agreement with density change measurements

Reactor physics and reactor computations

International Nuclear Information System (INIS)

Ronen, Y.; Elias, E.

1994-01-01

Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

A study of 2-Dimensional effects in the core of a PWR during the refloading phase of a LOCA. Analysis of data of PERICLES experiments with the COBRA-NC code

International Nuclear Information System (INIS)

Reinhardt, H.J.

1989-09-01

The project is embedded in the Shared Cost Action Programme (SCA) of the European Communities (CEC) on Reactor Safety, Research Area No. 4, concerning the analysis of experimental data on loss-of-coolant accidents and emergency core cooling. The PERICLES experiments, performed at CEA in Grenoble, had the objective to study multidimensional effects under well defined conditions concentrating on the inter-assembly character of reflood phenomena. The general aim of the present project is to analyse PERICLES experimental data in order to improve models in relevant system codes. Particular objectives of the project are - the critical evaluation of the experimental data of PERICLES Run 8; - the drawing of conclusions from the data with respect to physical and geometrical models for the multi-bundle reflood analysis; - the performance of one-and multi-dimensional computations with COBRA-NC; - the comparison of computational and experimental data; and - the development of conclusions and specifications of additional research needed. The analysis of the experimetal data of Run 8 was performed by a computer programme developed for postprocessing data of any PERICLES experiment. The postprocessor includes an automatic location of the quenchfront and displays it graphically with respect to time, vertical and horizontal directions. Furthermore, rod and fluid temperatures versus height, quenchtimes versus height, densities versus height, and temperatures, pressures, densities etc. versus time can be plotted. As far as computer simulations are concerned, it was one of the objectives of the present study to analyse in greater detail the multidimensional phenomena during the reflooding phase of a LOCA and to compare the numerical results with the experimental data. Such simulation may serve to adjust and improve existing computer codes as well as to validate the codes. Moreover, computer simulations are able to give information which are not available from experimental data; in the

Reactor

International Nuclear Information System (INIS)

Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

1979-01-01

Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Reactor container

International Nuclear Information System (INIS)

Kato, Masami; Nishio, Masahide.

1987-01-01

Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

Structural and functional characterization of a novel homodimeric three-finger neurotoxin from the venom of Ophiophagus hannah (king cobra).

Science.gov (United States)

Roy, Amrita; Zhou, Xingding; Chong, Ming Zhi; D'hoedt, Dieter; Foo, Chun Shin; Rajagopalan, Nandhakishore; Nirthanan, Selvanayagam; Bertrand, Daniel; Sivaraman, J; Kini, R Manjunatha

2010-03-12

Snake venoms are a mixture of pharmacologically active proteins and polypeptides that have led to the development of molecular probes and therapeutic agents. Here, we describe the structural and functional characterization of a novel neurotoxin, haditoxin, from the venom of Ophiophagus hannah (King cobra). Haditoxin exhibited novel pharmacology with antagonism toward muscle (alphabetagammadelta) and neuronal (alpha(7), alpha(3)beta(2), and alpha(4)beta(2)) nicotinic acetylcholine receptors (nAChRs) with highest affinity for alpha(7)-nAChRs. The high resolution (1.5 A) crystal structure revealed haditoxin to be a homodimer, like kappa-neurotoxins, which target neuronal alpha(3)beta(2)- and alpha(4)beta(2)-nAChRs. Interestingly however, the monomeric subunits of haditoxin were composed of a three-finger protein fold typical of curaremimetic short-chain alpha-neurotoxins. Biochemical studies confirmed that it existed as a non-covalent dimer species in solution. Its structural similarity to short-chain alpha-neurotoxins and kappa-neurotoxins notwithstanding, haditoxin exhibited unique blockade of alpha(7)-nAChRs (IC(50) 180 nm), which is recognized by neither short-chain alpha-neurotoxins nor kappa-neurotoxins. This is the first report of a dimeric short-chain alpha-neurotoxin interacting with neuronal alpha(7)-nAChRs as well as the first homodimeric three-finger toxin to interact with muscle nAChRs.

Joseph Clover and the cobra: a tale of snake envenomation and attempted resuscitation with bellows in London, 1852.

Science.gov (United States)

Ball, C

2010-07-01

The Industrial Revolution saw the creation of many new jobs, but probably none more curious than that of zookeeper. The London Zoological Gardens, established for members in 1828, was opened to the general public in 1847. In 1852 the "Head Keeper in the Serpent Room", Edward Horatio Girling, spent a night farewelling a friend departing for Australia. He arrived at work in an inebriated state and was bitten on the face by a cobra that he was handling in a less than sensible manner. He was taken by cab to University College Hospital where he was resuscitated by a number of doctors, including Joseph Clover then the resident medical officer to the hospital and later to become the leading anaesthetist in London. Clover recorded this event in his diary along with the resuscitation method used. The patient eventually died but his treatment created a flurry of correspondence in the medical and lay press. Interestingly, the attempted resuscitation was with bellows, which had been abandoned by the Royal Humane Society twenty years earlier Clover records other cases of resuscitation with bellows at University College Hospital during his time as a resident medical officer there (1848 to 1853). There is a casebook belonging to Joseph Clover in the Geoffrey Kaye Museum, in Melbourne. This story is one of the many interesting stories uncovered during a study of this book and Clover's other personal papers.

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Slurry reactors

Energy Technology Data Exchange (ETDEWEB)

Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

1979-08-01

Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Space-time reactor kinetics for heterogeneous reactor structure

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1969-11-15

An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

COBRA, an Arabidopsis extracellular glycosyl-phosphatidyl inositol-anchored protein, specifically controls highly anisotropic expansion through its involvement in cellulose microfibril orientation.

Science.gov (United States)

Roudier, François; Fernandez, Anita G; Fujita, Miki; Himmelspach, Regina; Borner, Georg H H; Schindelman, Gary; Song, Shuang; Baskin, Tobias I; Dupree, Paul; Wasteneys, Geoffrey O; Benfey, Philip N

2005-06-01

The orientation of cell expansion is a process at the heart of plant morphogenesis. Cellulose microfibrils are the primary anisotropic material in the cell wall and thus are likely to be the main determinant of the orientation of cell expansion. COBRA (COB) has been identified previously as a potential regulator of cellulose biogenesis. In this study, characterization of a null allele, cob-4, establishes the key role of COB in controlling anisotropic expansion in most developing organs. Quantitative polarized-light and field-emission scanning electron microscopy reveal that loss of anisotropic expansion in cob mutants is accompanied by disorganization of the orientation of cellulose microfibrils and subsequent reduction of crystalline cellulose. Analyses of the conditional cob-1 allele suggested that COB is primarily implicated in microfibril deposition during rapid elongation. Immunodetection analysis in elongating root cells revealed that, in agreement with its substitution by a glycosylphosphatidylinositol anchor, COB was polarly targeted to both the plasma membrane and the longitudinal cell walls and was distributed in a banding pattern perpendicular to the longitudinal axis via a microtubule-dependent mechanism. Our observations suggest that COB, through its involvement in cellulose microfibril orientation, is an essential factor in highly anisotropic expansion during plant morphogenesis.

RA Reactor

International Nuclear Information System (INIS)

1978-02-01

In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

H Reactor

Data.gov (United States)

Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

Diversion cross-flow mixing at the inlet of a simulated rod bundle using a gamma camera

International Nuclear Information System (INIS)

Sedaghat, A.; Macduff, R.; Castellana, F.

1986-01-01

The prediction of diversion cross-flow and turbulent mixing interests reactor vendors and nuclear fuel suppliers because of the effect on critical heat flux. In single-phase flow with uniform inlet conditions, flow diversion occurs primarily near the inlet. Prior work by Bowring and Levy and Lahey estimated diversion length by comparing the axial pressure differential at the channel exit using isokinetic (natural flow split) and nonisokinetic (forced flow split) sampling and by using a mathematical model. The present work, sponsored by Exxon Nuclear Company, Inc., represents the first study in which flow distribution and diversion cross flow were investigated at the inlet of a clean geometry. The parameters investigated were diversion length and the effective cross-flow velocity was determined by analysis. The results of this work were compared to theoretical values predicted by the COBRA IIIC subchannel computer code. The difference between experimental data and COBRA IIIC suggests that a more comprehensive transverse momentum balance is desired as mass flux ratios become large. The inclusion of transverse inertia and acceleration terms in the transverse momentum balance become important

Steady state subchannel analysis of AHWR fuel cluster

International Nuclear Information System (INIS)

Dasgupta, A.; Chandraker, D.K.; Vijayan, P.K.; Saha, D.

2006-09-01

Subchannel analysis is a technique used to predict the thermal hydraulic behavior of reactor fuel assemblies. The rod cluster is subdivided into a number of parallel interacting flow subchannels. The conservation equations are solved for each of these subchannels, taking into account subchannel interactions. Subchannel analysis of AHWR D-5 fuel cluster has been carried out to determine the variations in thermal hydraulic conditions of coolant and fuel temperatures along the length of the fuel bundle. The hottest regions within the AHWR fuel bundle have been identified. The effect of creep on the fuel performance has also been studied. MCHFR has been calculated using Jansen-Levy correlation. The calculations have been backed by sensitivity analysis for parameters whose values are not known accurately. The sensitivity analysis showed the calculations to have a very low sensitivity to these parameters. Apart from the analysis, the report also includes a brief introduction of a few subchannel codes. A brief description of the equations and solution methodology used in COBRA-IIIC and COBRA-IV-I is also given. (author)

Reactor container

International Nuclear Information System (INIS)

Naruse, Yoshihiro.

1990-01-01

The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

Hybrid reactors

International Nuclear Information System (INIS)

Moir, R.W.

1980-01-01

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

Multiregion reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

Reactor building

International Nuclear Information System (INIS)

Maruyama, Toru; Murata, Ritsuko.

1996-01-01

In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

Critical heat flux experiments for high conversion light water reactor, (3)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

1990-03-01

As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Reactor physics challenges in GEN-IV reactor design

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

2005-02-15

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

Reactor physics challenges in GEN-IV reactor design

International Nuclear Information System (INIS)

Driscoll, Michael K.; Hejzlar, Pavel

2005-01-01

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

Reactor Vessel Surveillance Program for Advanced Reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

2008-10-15

This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Molecular mechanism of cell death induced by king cobra (Ophiophagus hannah) venom l-amino acid oxidase.

Science.gov (United States)

Fung, Shin Yee; Lee, Mui Li; Tan, Nget Hong

2015-03-01

Snake venom LAAOs have been reported to exhibit a wide range of pharmacological activities, including cytotoxic, edema-inducing, platelet aggregation-inducing/platelet aggregation-inhibiting, bactericidal and antiviral activities. A heat-stable form of l-amino acid oxidase isolated from king cobra (Ophiophagus hannah) venom (OH-LAAO) has been shown to exhibit very potent cytotoxicity against human tumorigenic cells but not in their non-tumorigenic counterparts, and the cytotoxicity was due to the apoptosis-inducing effect of the enzyme. In this work, the molecular mechanism of cell death induced by OH-LAAO was investigated. The enzyme exerts its apoptosis-inducing effect presumably via both intrinsic and extrinsic pathways as suggested by the increase in caspase-8 and -9 activities. Oligonucleotide microarray analysis showed that the expression of a total of 178 genes was significantly altered as a result of oxidative stress induced by the hydrogen peroxide generated by the enzyme. Of the 178 genes, at least 27 genes are involved in apoptosis and cell death. These alterations of gene expression was presumably caused by the direct cytotoxic effect of H2O2 generated during the enzymatic reaction, as well as the non-specific oxidative modifications of signaling molecules that eventually lead to apoptosis and cell death. The very substantial up-regulation of cytochrome P450 genes may also contribute to the potent cytotoxic action of OH-LAAO by producing excessive reactive oxygen species (ROS). In conclusion, the potent apoptosis inducing activity of OH-LAAO was likely due to the direct cytotoxic effect of H2O2 generated during the enzymatic reaction, as well as the non-specific oxidation of signalling molecules. Copyright © 2015 Elsevier Ltd. All rights reserved.

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Nuclear Reactor Physics

Science.gov (United States)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

International Nuclear Information System (INIS)

Eustache, S.

1999-01-01

This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

Comparison of proteomic profiles of the venoms of two of the 'Big Four' snakes of India, the Indian cobra (Naja naja) and the common krait (Bungarus caeruleus), and analyses of their toxins.

Science.gov (United States)

Choudhury, Manisha; McCleary, Ryan J R; Kesherwani, Manish; Kini, R Manjunatha; Velmurugan, Devadasan

2017-09-01

Snake venoms are mixtures of biologically-active proteins and peptides, and several studies have described the characteristics of some of these toxins. However, complete proteomic profiling of the venoms of many snake species has not yet been done. The Indian cobra (Naja naja) and common krait (Bungarus caeruleus) are elapid snake species that are among the 'Big Four' responsible for the majority of human snake envenomation cases in India. As understanding the composition and complexity of venoms is necessary for successful treatment of envenomation in humans, we utilized three different proteomic profiling approaches to characterize these venoms: i) one-dimensional SDS-PAGE coupled with in-gel tryptic digestion and electrospray tandem mass spectrometry (ESI-LC-MS/MS) of individual protein bands; ii) in-solution tryptic digestion of crude venoms coupled with ESI-LC-MS/MS; and iii) separation by gel-filtration chromatography coupled with tryptic digestion and ESI-LC-MS/MS of separated fractions. From the generated data, 81 and 46 different proteins were identified from N. naja and B. caeruleus venoms, respectively, belonging to fifteen different protein families. Venoms from both species were found to contain a variety of phospholipases A 2 and three-finger toxins, whereas relatively higher numbers of snake venom metalloproteinases were found in N. naja compared to B. caeruleus venom. The analyses also identified less represented venom proteins including L-amino acid oxidases, cysteine-rich secretory proteins, 5'-nucleotidases and venom nerve growth factors. Further, Kunitz-type serine protease inhibitors, cobra venom factors, phosphodiesterases, vespryns and aminopeptidases were identified in the N. naja venom, while acetylcholinesterases and hyaluronidases were found in the B. caeruleus venom. We further analyzed protein coverage (Lys/Arg rich and poor regions as well as potential glycosylation sites) using in-house software. These studies expand our

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

FBR type reactors

International Nuclear Information System (INIS)

Suzuoki, Akira; Yamakawa, Masanori.

1985-01-01

Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

Nuclear reactor instrumentation at research reactor renewal

International Nuclear Information System (INIS)

Baers, B.; Pellionisz, P.

1981-10-01

The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

Reactor System Design

International Nuclear Information System (INIS)

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

2006-08-01

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

BWR type reactor

International Nuclear Information System (INIS)

Watanabe, Shoichi

1983-01-01

Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

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

Reactor core of nuclear reactor

International Nuclear Information System (INIS)

Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

1994-01-01

In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

High-temperature and breeder reactors - economic nuclear reactors of the future

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

Heterogeneous reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

Nuclear reactor neutron shielding

Science.gov (United States)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

2017-09-12

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

Fast breeder reactors

International Nuclear Information System (INIS)

Heinzel, V.

1975-01-01

The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

Reactor container

International Nuclear Information System (INIS)

Shibata, Satoru; Kawashima, Hiroaki

1984-01-01

Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

Reactor protection systems for the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Morris, C.R.

2003-01-01

The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

Reactor container

International Nuclear Information System (INIS)

Fukazawa, Masanori.

1991-01-01

A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

A multi-purpose reactor

International Nuclear Information System (INIS)

Changwen Ma

2000-01-01

An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Supercritical Water Reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

2014-01-01

The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

Tank type reactor

International Nuclear Information System (INIS)

Otsuka, Fumio.

1989-01-01

The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

New about research reactors

International Nuclear Information System (INIS)

Egorenkov, P.M.

2001-01-01

The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

International Nuclear Information System (INIS)

Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

2013-01-01

The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

Design guide for Category III reactors: pool type reactors

International Nuclear Information System (INIS)

Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

1978-11-01

The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems

Two acidic, anticoagulant PLA2 isoenzymes purified from the venom of monocled cobra Naja kaouthia exhibit different potency to inhibit thrombin and factor Xa via phospholipids independent, non-enzymatic mechanism.

Directory of Open Access Journals (Sweden)

Ashis K Mukherjee

Full Text Available The monocled cobra (Naja kaouthia is responsible for snakebite fatality in Indian subcontinent and in south-western China. Phospholipase A2 (PLA2; EC 3.1.1.4 is one of the toxic components of snake venom. The present study explores the mechanism and rationale(s for the differences in anticoagulant potency of two acidic PLA2 isoenzymes, Nk-PLA2α (13463.91 Da and Nk-PLA2β (13282.38 Da purified from the venom of N. kaouthia.By LC-MS/MS analysis, these PLA2s showed highest similarity (23.5% sequence coverage with PLA2 III isolated from monocled cobra venom. The catalytic activity of Nk-PLA2β exceeds that of Nk-PLA2α. Heparin differentially regulated the catalytic and anticoagulant activities of these Nk-PLA2 isoenzymes. The anticoagulant potency of Nk-PLA2α was comparable to commercial anticoagulants warfarin, and heparin/antithrombin-III albeit Nk-PLA2β demonstrated highest anticoagulant activity. The anticoagulant action of these PLA2s was partially contributed by a small but specific hydrolysis of plasma phospholipids. The strong anticoagulant effect of Nk-PLA2α and Nk-PLA2β was achieved via preferential, non-enzymatic inhibition of FXa (Ki = 43 nM and thrombin (Ki = 8.3 nM, respectively. Kinetics study suggests that the Nk-PLA2 isoenzymes inhibit their "pharmacological target(s" by uncompetitive mechanism without the requirement of phospholipids/Ca(2+. The anticoagulant potency of Nk-PLA2β which is higher than that of Nk-PLA2α is corroborated by its superior catalytic activity, its higher capacity for binding to phosphatidylcholine, and its greater strength of thrombin inhibition. These PLA2 isoenzymes thus have evolved to affect haemostasis by different mechanisms. The Nk-PLA2β partially inhibited the thrombin-induced aggregation of mammalian platelets suggesting its therapeutic application in the prevention of unwanted clot formation.In order to develop peptide-based superior anticoagulant therapeutics, future application of Nk-PLA2

The computerized reactor period measurement system for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1996-01-01

The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

Computer measurement system of reactor period for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1997-01-01

The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

Reactor safety method

International Nuclear Information System (INIS)

Vachon, L.J.

1980-01-01

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

Small propulsion reactor design based on particle bed reactor concept

International Nuclear Information System (INIS)

Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

1989-01-01

In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

Physics of nuclear reactors

International Nuclear Information System (INIS)

Baeten, Peter

2006-01-01

This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

Next-generation genome-scale models for metabolic engineering

DEFF Research Database (Denmark)

King, Zachary A.; Lloyd, Colton J.; Feist, Adam M.

2015-01-01

Constraint-based reconstruction and analysis (COBRA) methods have become widely used tools for metabolic engineering in both academic and industrial laboratories. By employing a genome-scale in silico representation of the metabolic network of a host organism, COBRA methods can be used to predict...... examples of applying COBRA methods to strain optimization are presented and discussed. Then, an outlook is provided on the next generation of COBRA models and the new types of predictions they will enable for systems metabolic engineering....

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

Guide to power reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-07-15

The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

Analysis of dynamic stability and safety of reactor system by reactor simulator

International Nuclear Information System (INIS)

Raisic, N.

1963-11-01

In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

Nuclear reactor physics

CERN Document Server

Stacey, Weston M

2010-01-01

Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

Atomic reactor thermal engineering

International Nuclear Information System (INIS)

Kim, Gwang Ryong

1983-02-01

This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

Argentinean integrated small reactor design and scale economy analysis of integrated reactor

International Nuclear Information System (INIS)

Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

2000-01-01

This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

Development of a nuclear power plant system analysis code

International Nuclear Information System (INIS)

Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

1997-07-01

During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

Fast reactors

International Nuclear Information System (INIS)

Vasile, A.

2001-01-01

Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

Reactor power measuring device

International Nuclear Information System (INIS)

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

1993-01-01

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

Reactor container

International Nuclear Information System (INIS)

Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

1997-01-01

In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

To the analysis of reactor noise in boiling water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1972-01-01

The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

The CEA research reactors

International Nuclear Information System (INIS)

Schwartz, J.P.

1993-01-01

Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Virtual nuclear reactor for education of nuclear reactor physics

International Nuclear Information System (INIS)

Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

2008-01-01

As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

International Nuclear Information System (INIS)

Schikorr, W.M.

In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

Reactor core

International Nuclear Information System (INIS)

Azekura, Kazuo; Kurihara, Kunitoshi.

1992-01-01

In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

Nuclear reactor

International Nuclear Information System (INIS)

Batheja, P.; Huber, R.; Rau, P.

1985-01-01

Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

Energy Technology Data Exchange (ETDEWEB)

Kliem, S.

2017-09-15

The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

The fast breeder reactor

International Nuclear Information System (INIS)

Collier, J.

1990-01-01

The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

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

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Ulysse, mentor reactor

International Nuclear Information System (INIS)

Bouquin, B.; Rio, I.; Safieh, J.

1997-01-01

On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Experience in using a research reactor for the training of power reactor operators

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenaut, L.J.

1972-01-01

A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

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

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

Science.gov (United States)

Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

2016-01-01

Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

The fast reactor

International Nuclear Information System (INIS)

1980-02-01

The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Masood, Z.

2016-01-01

The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

Fast Spectrum Reactors

CERN Document Server

Todd, Donald; Tsvetkov, Pavel

2012-01-01

Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1985-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

1982-09-01

Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

International Nuclear Information System (INIS)

Shiba, T.; Fallot, M.

2015-01-01

To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

Reactor Sharing Program

International Nuclear Information System (INIS)

Tehan, Terry

2002-01-01

Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

NUCLEAR REACTOR

Science.gov (United States)

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Reactor control device

International Nuclear Information System (INIS)

Fukami, Haruo; Morimoto, Yoshinori.

1981-01-01

Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

Research reactor DHRUVA

International Nuclear Information System (INIS)

Veeraraghaven, N.

1990-01-01

DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

Development of sub-channel/system coupled code and its application to a supercritical water-cooled test loop

International Nuclear Information System (INIS)

Liu, X.J.; Yang, T.; Cheng, X.

2014-01-01

To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code and system code are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal-hydraulic parameters are predicted by the sub-channel code COBRA-SC. Sensitivity analysis are carried out respectively in ATHLET-SC and COBRA-SC code, to identify the appropriate models for description of the flow blockage phenomenon in the test loop. Some measures to mitigate the accident consequence are also trialed to demonstrate their effectiveness. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel assembly can be reduced effectively by the safety measures of SCWR-FQT. (author)

Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

International Nuclear Information System (INIS)

Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

1992-01-01

A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

Reactor safety device

International Nuclear Information System (INIS)

Okada, Yasumasa.

1987-01-01

Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1980-09-01

Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

Reactor scram device for FBR type reactor

International Nuclear Information System (INIS)

Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

1991-01-01

In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

Refuelling nuclear reactors

International Nuclear Information System (INIS)

Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

1981-01-01

An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

Compilation of reactor physics data of the year 1984, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-12-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1984 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

Compilation of reactor physics data of the year 1983, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-06-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1983 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

Reactor containment

International Nuclear Information System (INIS)

Kawabe, Ryuhei; Yamaki, Rika.

1990-01-01

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

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1978-10-01

Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

NARCIS (Netherlands)

Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

2009-01-01

In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

Undergraduate reactor control experiment

International Nuclear Information System (INIS)

Edwards, R.M.; Power, M.A.; Bryan, M.

1992-01-01

A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

Advances in reactor physics education: Visualization of reactor parameters

International Nuclear Information System (INIS)

Snoj, L.; Kromar, M.; Zerovnik, G.

2012-01-01

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

Reactor feedwater system

International Nuclear Information System (INIS)

Hikabe, Katsumi.

1978-01-01

Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

New reactor concepts

International Nuclear Information System (INIS)

Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

1998-11-01

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

Fission reactors and materials

International Nuclear Information System (INIS)

Frost, B.R.T.

1981-12-01

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J. Jr.

1981-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

Method of operating a reactor

International Nuclear Information System (INIS)

Oosumi, Katsumi; Yamamoto, Michiyoshi.

1980-01-01

Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

Reactor container

International Nuclear Information System (INIS)

Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

1987-01-01

Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1993-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor engineering department annual report

International Nuclear Information System (INIS)

1990-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Belgia modernistid kunstihoones

Index Scriptorium Estoniae

2010-01-01

Retrospektiivnäitus Belgia modernistliku maalikunsti ajaloost "CoBrA & Co" Tallinna Kunstihoones 7.12.2010-21.01.2011, kuraatorid Denis Laoureux (Belgia) ja Harry Liivrand. Näitusel antakse ülevaade ekspressionistliku rühmituse CoBrA (1948-1951) maaliloomingust ja demonstreeritakse CoBrA eeskujudele toetuvat ja tema pärandit interpreteerivat kunstipraktikat Valloonias

Status of work at PNL supporting dry storage of spent fuel

International Nuclear Information System (INIS)

Cunningham, M.E.; McKinnon, M.A.; Michener, T.E.; Thomas, L.E.; Thornhill, C.K.

1992-01-01

Three projects related to dry storage of light-water reactor spent fuel are being conducted at Pacific Northwest Laboratory. Performance testing of six dry storage systems (four metal casks and two concrete storage systems) has been completed and results compiled. Two computer codes for predicting spent fuel and storage system thermal performance, COBRA-SFS and HYDRA-II, have been developed and have been reviewed by the US Nuclear Regulatory Commission. Air oxidation testing of spent fuel was conducted from 1984 through 1990 to obtain data to support recommendations of temperature-time limits for air dry storage for periods up to 40 years

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

1977-09-01

Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1984-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1976-09-01

Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

Department of reactor technology

International Nuclear Information System (INIS)

1980-01-01

The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Remote Reactor Monitoring

Energy Technology Data Exchange (ETDEWEB)

Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-10-21

The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

2001-07-01

Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

Canada-India Reactor (CIR)

Energy Technology Data Exchange (ETDEWEB)

None

1960-12-15

Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

International Nuclear Information System (INIS)

Nelson, W.R.

1982-01-01

REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

One piece reactor removal

International Nuclear Information System (INIS)

Chia, Wei-Min; Wang, Song-Feng

1993-01-01

The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

Nuclear reactors

International Nuclear Information System (INIS)

Middleton, J.E.

1977-01-01

Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

Reactor decommissioning

International Nuclear Information System (INIS)

Lawton, H.

1984-01-01

A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

Effects of snake venom from Saudi cobras and vipers on hormonal levels in peripheral blood.

Science.gov (United States)

Abdel-Galil, Khidir A; Al-Hazimi, Awdah M

2004-08-01

Knowledge about the effects of snake venoms on endocrine glands in the Kingdom of Saudi Arabia (KSA) is meager. The aim of the present study is to investigate the acute and chronic envenomation from 4 snakes out of 8 species of Saudi Cobras and Vipers on the tissues of endocrine glands and peripheral hormonal levels in male rats. The peripheral blood levels of 4 hormones mainly testosterone, cortisol, insulin and thyroxin were investigated in male Wistar rats following acute and chronic treatment of the rats with poisonous snake venoms at the Department of Physiology, Faculty of Medicine, King Abdul-Aziz University, Jeddah, Kingdom of Saudi Arabia between September 2000 to May 2001. Using radio immunoassay for hormonal analysis, a rise in testosterone levels in peripheral blood was obtained following acute treatment, which is due to the effect of the venoms on vascular permeability and increased blood flow. In contrast, the chronic treatment with venoms resulted in a delayed effect on vascular permeability and testicular degeneration resulting in a decreased blood flow and a significant drop in testosterone concentration. Cortisol levels were no different from the controls during acute treatment but it demonstrates gradual rise following chronic treatment to withstand the stress imposed on the animals. Similar results were obtained for insulin, which showed normal values with acute treatment but decreased levels of chronic treatment suggesting insulin insufficiently. Likewise, the thyroxin levels were decreased with chronic treatment suggesting a toxic effect of the poison on the rich blood supply of the thyroid follicles with a subsequent decrease in blood flow to the tissues and therefore, decreased thyroid hormone levels. The effects of venom toxicity on testosterone levels were either normal or stimulatory with acute treatment or inhibitory with chronic treatment depending on the vascular blood flow and testicular degeneration. Cortisol levels were normal at

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01