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Sample records for paks-3 reactor

  1. Reactor protection system refurbishment at Paks

    International Nuclear Information System (INIS)

    Hetzmann, A.; Turi, T.

    1997-01-01

    The history and the milestones of the reactor protection system refurbishment are outlined. During the preparation phase of the refurbishment project, detailed requirements have been set up and specific technical solutions developed. The structure of the project documents prepared during these activities is shown in a figure. The life cycle of the project was divided into four phases: the preparatory phase; the design and manufacturing phase; the installation and commissioning phase; and the operation phase. For all four Paks units a time schedule for implementation was set up. The licensing process is dealt with; the principal license was issued in June 1996. (A.K.)

  2. Integral tightness measurements at the Paks-1 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taubner, R.; Techy, Z. (Villamosenergiaipari Kutato Intezet, Budapest (Hungary))

    1983-01-01

    The containment system experiments of the Paks-1 nuclear reactor are described. The integrated tightness measurements of the hermetic system were completed in 1982. The principles and methods and the evaluation of the results of the measurements are discussed. Some features of the filtration characteristics are demonstrated using relative values and a method enabling the description of the physical contents of the characteristics by flow technical functions is outlined.

  3. PAK6 Phosphorylates 14-3-3γ to Regulate Steady State Phosphorylation of LRRK2

    Directory of Open Access Journals (Sweden)

    Laura Civiero

    2017-12-01

    Full Text Available Mutations in Leucine-rich repeat kinase 2 (LRRK2 are associated with Parkinson's disease (PD and, as such, LRRK2 is considered a promising therapeutic target for age-related neurodegeneration. Although the cellular functions of LRRK2 in health and disease are incompletely understood, robust evidence indicates that PD-associated mutations alter LRRK2 kinase and GTPase activities with consequent deregulation of the downstream signaling pathways. We have previously demonstrated that one LRRK2 binding partner is P21 (RAC1 Activated Kinase 6 (PAK6. Here, we interrogate the PAK6 interactome and find that PAK6 binds a subset of 14-3-3 proteins in a kinase dependent manner. Furthermore, PAK6 efficiently phosphorylates 14-3-3γ at Ser59 and this phosphorylation serves as a switch to dissociate the chaperone from client proteins including LRRK2, a well-established 14-3-3 binding partner. We found that 14-3-3γ phosphorylated by PAK6 is no longer competent to bind LRRK2 at phospho-Ser935, causing LRRK2 dephosphorylation. To address whether these interactions are relevant in a neuronal context, we demonstrate that a constitutively active form of PAK6 rescues the G2019S LRRK2-associated neurite shortening through phosphorylation of 14-3-3γ. Our results identify PAK6 as the kinase for 14-3-3γ and reveal a novel regulatory mechanism of 14-3-3/LRRK2 complex in the brain.

  4. Upgrading the reactor noise diagnostic systems at the Paks NPP

    International Nuclear Information System (INIS)

    Czibok, T.; Dezsoe, Z.; Kiss, K.; Krinizs, K.; Lipcsei, S.

    2002-01-01

    The paper reports on the actual step in upgrading process of the reactor noise diagnostic systems at Paks NPP. This step has mainly a technical character. Renewal of facilities for signal conditioning and for data acquisition is going on. Autonomous systems at each of the four reactor units will be able to acquire a set of data series which can be arbitrarily chosen from the whole set of several hundred in-core neutron and other signals. The autonomous systems can be remotely controlled by a central computer through the local network. Modularity and extensibility are important features of the new systems: the size of the set of available signals can be extended and new modules for more advanced evaluations can be installed later. Present plans for system hardware upgrading are outlined, together with some technical details of measurement control, data acquisition moduls and network communication.(abstract)

  5. Refurbishment of the reactor protection system at Paks NPP. The refurbishment process

    International Nuclear Information System (INIS)

    Turi, T.; Katics, B.

    1998-01-01

    The Reactor Protection System Refurbishment Project had an extensive preparation period in Paks started in 1992. During this preparation a large volume of the basic engineering tasks has been performed and as a result a contract for implementation of a three-train digital RPS on the four Units was concluded with Siemens in September, 1996. According to that contract the first refurbished Unit will be commissioned in 1999 followed by a further Unit in each succeeding year. This paper introduces the process of the refurbishment, overview of the V and V activities, introduce the architecture, summarize the main design principles and outlines the additional tasks to be performed together with the RPS design. (author)

  6. Reactor safety instrumentation of Paks NPP (experience and perspective)

    International Nuclear Information System (INIS)

    Elo, S.; Hamar, K.

    1993-01-01

    The majority of the existing control and protection systems in nuclear power plants use old analog technology and design philosophy. Maintenance and the procurement of spare parts is becoming increasingly difficult. In general there is an age degradation concern. Aging degradation in nuclear power plants must be effectively managed to avoid a loss of vital safety function, shutdown of the station, a reduced power generation, or any failure leading to expensive repair. Even with the best efforts in developing reliable and long life instrumentation and control systems for nuclear power plants it is expected that these systems for most plants will require replacements during the life of the plants. The instrumentation and control system of the nuclear power plants designed during the 70's and constructed in the 80's went out-of-date since nuclear safety is not a static concept and the digital computer technology has undergone rapid improvements during the 70's and 80's. Simultaneously the operation and the maintenance of the I ampersand C system of those plants described above becomes more and more difficult and expensive. In this context the pure quality of the former Soviet designed process instrumentation system increases the needs of upgrading this system. The author reviews the main design characteristics of the reactor safety instrumentation of the Paks NPP. Further he attempts to convey the perspective on upgrading the reactor safety instrumentation as seen by the HAEC and its Nuclear Safety Inspectorate

  7. Functional PAK-2 knockout and replacement with a caspase cleavage-deficient mutant in mice reveals differential requirements of full-length PAK-2 and caspase-activated PAK-2p34.

    Science.gov (United States)

    Marlin, Jerry W; Chang, Yu-Wen E; Ober, Margaret; Handy, Amy; Xu, Wenhao; Jakobi, Rolf

    2011-06-01

    p21-Activated protein kinase 2 (PAK-2) has both anti- and pro-apoptotic functions depending on its mechanism of activation. Activation of full-length PAK-2 by the monomeric GTPases Cdc42 or Rac stimulates cell survival, whereas caspase activation of PAK-2 to the PAK-2p34 fragment is involved in the apoptotic response. In this study we use functional knockout of PAK-2 and gene replacement with the caspase cleavage-deficient PAK-2D212N mutant to differentiate the biological functions of full-length PAK-2 and caspase-activated PAK-2p34. Knockout of PAK-2 results in embryonic lethality at early stages before organ development, whereas replacement with the caspase cleavage-deficient PAK-2D212N results in viable and healthy mice, indicating that early embryonic lethality is caused by deficiency of full-length PAK-2 rather than lack of caspase activation to the PAK-2p34 fragment. However, deficiency of caspase activation of PAK-2 decreased spontaneous cell death of primary mouse embryonic fibroblasts and increased cell growth at high cell density. In contrast, stress-induced cell death by treatment with the anti-cancer drug cisplatin was not reduced by deficiency of caspase activation of PAK-2, but switched from an apoptotic to a nonapoptotic, caspase-independent mechanism. Homozygous PAK-2D212N primary mouse embryonic fibroblasts that lack the ability to generate the proapoptotic PAK-2p34 show less activation of the effector caspase 3, 6, and 7, indicating that caspase activation of PAK-2 amplifies the apoptotic response through a positive feedback loop resulting in more activation of effector caspases.

  8. Summary of structural analysis and comparison with experimental results for Paks NPP

    International Nuclear Information System (INIS)

    Hauptenbuchner, B.; David, M.

    2001-01-01

    This contribution deals with the analysis and comparison of the dynamic response, calculated and measured by the explosion test in Nuclear Power Plant Paks, Hungary. Some details of the calculation model are also presented. The calculated and measured data of dynamic response are compared in selected points of the NPP Paks reactor building. Conclusions and recommendations are derived from this comparison. (author)

  9. Reactor Dosimetry Aspects of the Service Life Extension of the Hungarian Paks NPP

    Directory of Open Access Journals (Sweden)

    Zsolnay Eva M.

    2016-01-01

    Full Text Available The service life of the Hungarian Paks Nuclear Power Plant (NPP will be extended from the originally planned 30 years to 50 years. To improve the reliability of the results obtained in frame of the old reactor pressure vessel (RPV surveillance programme, new methods have been developed, and based on them, the old exposition data have been re-evaluated for all the four reactor units. At the same time, a new RPV surveillance programme has been developed and introduced, and long term irradiations have been performed to determine the radiation damage of the surveillance specimens due to the high fast neutron exposition. Neutron transport calculations have been performed with a validated neutron transport code system to determine the fast neutron exposition of the RPVs during the extended service life. The cavity dosimetry is in the introductory phase. This paper presents the new developments in the field of the RPV surveillance dosimetry and summarises the results obtained. According to the results the service life of the NPP can safely be extended for the planned 50 years.

  10. Safety reassessment of the Paks NPP (the AGNES project)

    Energy Technology Data Exchange (ETDEWEB)

    Gado, J [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics; Bajsz, J; Cserhati, A; Elter, J [Paksi Atomeroemue Vallalat, Paks (Hungary); Hollo, E [Energiagazdalkodasi Intezet, Budapest (Hungary); Kovacs, K [EROTERV Engineering and Contractor Co (Hungary); Maroti, L [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics; Miko, S [Paksi Atomeroemue Vallalat, Paks (Hungary); Techy, Z [Energiagazdalkodasi Intezet, Budapest (Hungary); Vidovszky, I [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics

    1996-12-31

    The reassessment of the Paks NPP safety according to internationally recognized criteria of the Advanced General and New Evaluation of Safety (AGNES) project is outlined. The Paks NPP consists of four WWER-440/V-213 units. The following groups of analysis have been performed: system analysis and description; analysis of design basis accidents; severe accidents analysis; level 1 probabilistic safety analysis. Postulated accidents (PA) and Anticipated Operational Occurrences (AOO) are estimated in detail for the following initiating events: increase/decrease in secondary heat removal; decrease in primary coolant inventory; increase/decrease of reactor coolant inventory; reactivity and power distribution anomalies; analysis of transients with the failure of reactor scram (ATWS); pressurized thermal shock analyses. Severe accident analysis was made for the accidents on in-vessel phase and containment phase, for radioactive release and for accident management.

  11. Safety reassessment of the Paks NPP (the AGNES project)

    International Nuclear Information System (INIS)

    Gado, J.; Hollo, E.; Kovacs, K.; Maroti, L.; Techy, Z.; Vidovszky, I.

    1995-01-01

    The reassessment of the Paks NPP safety according to internationally recognized criteria of the Advanced General and New Evaluation of Safety (AGNES) project is outlined. The Paks NPP consists of four WWER-440/V-213 units. The following groups of analysis have been performed: system analysis and description; analysis of design basis accidents; severe accidents analysis; level 1 probabilistic safety analysis. Postulated accidents (PA) and Anticipated Operational Occurrences (AOO) are estimated in detail for the following initiating events: increase/decrease in secondary heat removal; decrease in primary coolant inventory; increase/decrease of reactor coolant inventory; reactivity and power distribution anomalies; analysis of transients with the failure of reactor scram (ATWS); pressurized thermal shock analyses. Severe accident analysis was made for the accidents on in-vessel phase and containment phase, for radioactive release and for accident management

  12. Extension of the surveillance program at NPP Paks

    International Nuclear Information System (INIS)

    Gillemot, F.

    1992-01-01

    In WWER-440 reactors the surveillance specimens are located in accelerated irradiation positions. After five years all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension is started to eliminate of this disadvantage of the original program. (author)

  13. Safer nuclear power. Strengthening training for operational safety at Paks nuclear power plant - Hungary

    International Nuclear Information System (INIS)

    2003-01-01

    For a nuclear power plant, safety must always be paramount. There can be no compromise on safety to meet production targets or to reduce costs. For any reactor, and in particular where older type reactors are in place, their operational safety can be enhanced by upgrading the training of personnel responsible for operating and maintaining the plant. The Department of Technical Co-operation is sponsoring a programme with technical support from the Nuclear Energy and Nuclear Safety Departments to help improve facilities at the PAKS plant in Hungary and establish self sufficiency in training to the highest international standards for all levels of nuclear power plant manpower. The Model Project described will have a direct impact on the improvement of operational safety and performance at PAKS NPP. It will lead to a more efficient use of resources which in turn will result in lower electricity generation costs. The impact of the project is not expected to be limited to Hungary. WWER reactors are common in Eastern Europe and provide one third to one half of the electricity supply to the region. The training programmes and facilities at PAKS offer a possibility in the future to provide training to experts from other countries operating WWER units and serve as a model to be emulated. Slovakia and the Czech Republic have already expressed interest in using the PAKS experience

  14. PAK1 translocates into nucleus in response to prolactin but not to estrogen

    Energy Technology Data Exchange (ETDEWEB)

    Oladimeji, Peter, E-mail: Peter.Oladimeji@rockets.utoledo.edu; Diakonova, Maria, E-mail: mdiakon@utnet.utoledo.edu

    2016-04-22

    Tyrosyl phosphorylation of the p21-activated serine–threonine kinase 1 (PAK1) has an essential role in regulating PAK1 functions in breast cancer cells. We previously demonstrated that PAK1 serves as a common node for estrogen (E2)- and prolactin (PRL)-dependent pathways. We hypothesize herein that intracellular localization of PAK1 is affected by PRL and E2 treatments differently. We demonstrate by immunocytochemical analysis that PAK1 nuclear translocation is ligand-dependent: only PRL but not E2 stimulated PAK1 nuclear translocation. Tyrosyl phosphorylation of PAK1 is essential for this nuclear translocation because phospho-tyrosyl-deficient PAK1 Y3F mutant is retained in the cytoplasm in response to PRL. We confirmed these data by Western blot analysis of subcellular fractions. In 30 min of PRL treatment, only 48% of pTyr-PAK1 is retained in the cytoplasm of PAK1 WT clone while 52% re-distributes into the nucleus and pTyr-PAK1 shuttles back to the cytoplasm by 60 min of PRL treatment. In contrast, PAK1 Y3F is retained in the cytoplasm. E2 treatment causes nuclear translocation of neither PAK1 WT nor PAK1 Y3F. Finally, we show by an in vitro kinase assay that PRL but not E2 stimulates PAK1 kinase activity in the nuclear fraction. Thus, PAK1 nuclear translocation is ligand-dependent: PRL activates PAK1 and induces translocation of activated pTyr-PAK1 into nucleus while E2 activates pTyr-PAK1 only in the cytoplasm. - Highlights: • Prolactin but not estrogen causes translocation of PAK1 into nucleus. • Tyrosyl phosphorylation of PAK1 is required for nuclear localization. • Prolactin but not estrogen stimulates PAK1 kinase activity in nucleus.

  15. Paks shows the way towards good operating practices

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    The Paks-3 unit in Hungary was the first VVER (Soviet designed Pressurized Water Reactor) to be scrutinized by an International Atomic Energy Agency Operational Safety Analysis Review Team. A number of examples of good operational practice were noted. Those reported here include the cleanliness of the plant, the management attitude to training, early detection of and action to correct problems as they arise, an accident avoidance policy, a back-up research and development programme, and the provision of computer-based assistance to the operator to present operational data in an easily comprehensible form. (U.K.)

  16. Vibration system identification of Paks and Kozloduy reactor buildings on the basis of the blast test results

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1999-01-01

    System identification allows to build mathematical models of a dynamic system based on measured data. System identification is carried out by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The aim of this study is to investigate and model the behavior of complex vibratory systems on the basis of measured excitation and response. The first part of the study describes the theory used in the analysis and the software tools used in the analysis. The second part of the study describes the investigation and modeling of the response of single degree of freedom oscillator excited by sinusoidal and blast excitation. In the third part of the study the system identification of the Kozloduy NPP unit 5 reactor building and Paks NPP unit 1 reactor building is studied and the models are estimated using the method of segmentation of excitation and response. System identification is carried out using MATLAB software by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The types of models used for the were: l) ARX models; 2) ARMAX model; 3) Output-Error (OE) models; 4) Box-Jenkins (BJ) models; 5) State-space models. The model coefficients for different models were calculated using the least-squares and maximum likelihood estimation methods available in MATLAB system identification toolbox. Excitation was in both Paks and Kozloduy case the measured free-field excitation and responses were the vibration responses of the building on the foundation slab level and top of the building. By examining the established models the frequency characteristics of vibration systems were determined with 95 % accuracy and the amplitude response with 80 % accuracy. In case of the steady state response of sinusoidally excited single dof oscillator the modelling gave almost exact results. But in the case of the blast response of the reactor building the obtaining of the

  17. Review of Paks outage results 1990

    International Nuclear Information System (INIS)

    Lukacs, P.; Zsoldos, F.; Kiss, Z.

    1991-01-01

    The year 1990 was not the most successful from an outage point of view at the Paks Nuclear Power Plant in Hungary -there were one or two long delays. Work at unit 4 had a delay of 10 days because of an error made during assembling the reactor vessel. While the outage of unit 3 was running, a feedwater pipe hanger problem was discovered - several hangers were found displaced from the right position. A general inspection of the affected system was required and this took about 11 days. Information about each outage is presented on diagrams, making comparison easier. These diagrams give information about deviations from the outage plan, about work hours performed during outages, and about collective exposure. (author)

  18. Nuclear safety at the Paks Plant

    International Nuclear Information System (INIS)

    Bajsz, Jozsef; Vamos, Gabor

    1991-01-01

    The Paks Nuclear Power Plant is located on the Danube river 114 km south of Budapest. It consists of four PWR units of the Soviet VVER-440 design. These are of the second generation design VVER 440 (model 213) with safety features as of 1975. It should be emphasized that these are two different generations of VVER 440 units. This is not always clear, not only to the general public, but sometimes even to people working in the nuclear industry. The widespread criticism of the first generation type 230 reactors is often extended to model 213 reactors, as the differences between the two models are often not sufficiently emphasized. In this situation it is very important to provide balanced information about the advantages and disadvantages of this reactor type. This paper attempts to do that. (author)

  19. The Paks Nuclear Power Station

    International Nuclear Information System (INIS)

    Erdosi, N.; Szabo, L.

    1978-01-01

    As the first stage in the construction of the Paks Nuclear Power Station, two units of 440 MW(e) each will be built. They are operated with two coolant loops each. The reactor units are VVER 440 type water-moderated PWR type heterogeneous power reactors designed in the Soviet Union and manufactured in Czechoslovakia. Each unit operates two Soviet-made K-220-44 steam turbines and Hungarian-made generators of an effective output of 220 MW. The output of the transformer units - also of Hungarian made - is 270 MVA. The radiation protection system of the nuclear power station is described. Protection against system failures is accomplished by specially designed equipment and security measures especially within the primary circuit. Some data on the power station under construction are given. (R.P.)

  20. Theses on fundamental issues of Paks extension debate in Hungary

    International Nuclear Information System (INIS)

    Ujhelyi, Geza

    2014-01-01

    The paper analyzes the expected price of electric power generated by the new blocks of the Paks Nuclear Power Plant. Takes into account the investment, the loan repayment, the depreciation period and the nuclear fuel prices. Calls the readers attention to the closing down of the existing four reactor blocks when reaching the end of the extended lifetime between 2032 and 2037. Compares the parameters of electric power generated by nuclear reactors with those of renewables. (TRA)

  1. Pak3 promotes cell cycle exit and differentiation of β-cells in the embryonic pancreas and is necessary to maintain glucose homeostasis in adult mice.

    Science.gov (United States)

    Piccand, Julie; Meunier, Aline; Merle, Carole; Jia, Zhengping; Barnier, Jean-Vianney; Gradwohl, Gérard

    2014-01-01

    The transcription factor neurogenin3 (Ngn3) triggers islet cell differentiation in the developing pancreas. However, little is known about the molecular mechanisms coupling cell cycle exit and differentiation in Ngn3(+) islet progenitors. We identified a novel effector of Ngn3 endocrinogenic function, the p21 protein-activated kinase Pak3, known to control neuronal differentiation and implicated in X-linked intellectual disability in humans. We show that Pak3 expression is initiated in Ngn3(+) endocrine progenitor cells and next maintained in maturing hormone-expressing cells during pancreas development as well as in adult islet cells. In Pak3-deficient embryos, the proliferation of Ngn3(+) progenitors and β-cells is transiently increased concomitantly with an upregulation of Ccnd1. β-Cell differentiation is impaired at E15.5 but resumes at later stages. Pak3-deficient mice do not develop overt diabetes but are glucose intolerant under high-fat diet (HFD). In the intestine, Pak3 is expressed in enteroendocrine cells but is not necessary for their differentiation. Our results indicate that Pak3 is a novel regulator of β-cell differentiation and function. Pak3 acts downstream of Ngn3 to promote cell cycle exit and differentiation in the embryo by a mechanism that might involve repression of Ccnd1. In the adult, Pak3 is required for the proper control of glucose homeostasis under challenging HFD.

  2. VERONA V6.22 – An enhanced reactor analysis tool applied for continuous core parameter monitoring at Paks NPP

    Energy Technology Data Exchange (ETDEWEB)

    Végh, J., E-mail: janos.vegh@ec.europa.eu [Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Pós, I., E-mail: pos@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Horváth, Cs., E-mail: csaba.horvath@energia.mta.hu [Centre for Energy Research, Hungarian Academy of Sciences, H-1525 Budapest 114, P.O. Box 49 (Hungary); Kálya, Z., E-mail: kalyaz@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Parkó, T., E-mail: parkot@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Ignits, M., E-mail: ignits@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary)

    2015-10-15

    Between 2003 and 2007 the Hungarian Paks NPP performed a large modernization project to upgrade its VERONA core monitoring system. The modernization work resulted in a state-of-the-art system that was able to support the reactor thermal power increase to 108% by more accurate and more frequent core analysis. Details of the new system are given in Végh et al. (2008), the most important improvements were as follows: complete replacement of the hardware and the local area network; application of a new operating system and porting a large fraction of the original application software to the new environment; implementation of a new human-system interface; and last but not least, introduction of new reactor physics calculations. Basic novelty of the modernized core analysis was the introduction of an on-line core-follow module based on the standard Paks NPP core design code HELIOS/C-PORCA. New calculations also provided much finer spatial resolution, both in terms of axial node numbers and within the fuel assemblies. The new system was able to calculate the fuel applied during the first phase of power increase accurately, but it was not tailored to determine the effects of burnable absorbers as gadolinium. However, in the second phase of the power increase process the application of fuel assemblies containing three fuel rods with gadolinium content was intended (in order to optimize fuel economy), therefore off-line and on-line VERONA reactor physics models had to be further modified, to be able to handle the new fuel according to the accuracy requirements. In the present paper first a brief overview of the system version (V6.0) commissioned after the first modernization step is outlined; then details of the modified off-line and on-line reactor physics calculations are described. Validation results for new modules are treated extensively, in order to illustrate the extent and complexity of the V&V procedure associated with the development and licensing of the new

  3. 3ON PAK RUPEE EXCHANGE RATES: WHETHER STOCK OR FLOW MATTERS?

    Directory of Open Access Journals (Sweden)

    Razzaque H Bhatti

    2011-01-01

    Full Text Available This paper examines whether the monetary model or the flow model of exchange rate explains the long-run movements in Pak rupee exchange rates vis-à-vis the four major currencies – the US dollar, British pound, Swiss franc and Japanese yen – over the period 1983q1-2009q4. Results obtained by employing the Johansen and Juselius (1990 technique of cointegration are supportive of the monetary model in two Pak rupee exchange rates vis-à-vis the US dollar and the Swiss franc when both short- and long-run interest rates are used and of the flow model in three exchange rates vis-à-vis the British pound, Swiss franc and Japanese yen when the short-run interest rate is used. These results show that both stock equilibrium in capital markets and flow equilibrium in foreign exchange markets determine Pak rupee exchange rates.

  4. Refuelling design and core calculations at NPP Paks: codes and methods

    International Nuclear Information System (INIS)

    Pos, I.; Nemes, I.; Javor, E.; Korpas, L.; Szecsenyi, Z.; Patai-Szabo, S.

    2001-01-01

    This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)

  5. OECD-IAEA Paks Fuel Project. Final Report

    International Nuclear Information System (INIS)

    2010-05-01

    It is important for nuclear power plant designers, operators and regulators to effectively use lessons learned from events occurring at nuclear power plants since, in general, it is impossible to reproduce the event using experimental facilities. In particular, evaluation of the event using accident analysis codes is expected to contribute to improving understanding of phenomena during the events and to facilitate the validation of computer codes through simulation analyses. The information presented in this publication will be of use in future revisions of safety guides on accident analysis. During a fuel crud removal operation on the Paks-2 unit of the Paks nuclear power plant, Hungary on 10 April 2003, several fuel assemblies were severely damaged. The assemblies were being cleaned in a special tank under deep water in a service pit connected to the spent fuel storage pool. The first sign of fuel failures was the detection of some fission gases released from the cleaning tank. Later, visual inspection revealed that most of the 30 fuel assemblies suffered heavy oxidation and fragmentation. The first evaluation of the event showed that the severe fuel damage had been caused by inadequate cooling. The Paks-2 event was discussed in various committees of the OECD Nuclear Energy Agency (OECD/NEA) and of the International Atomic Energy Agency (IAEA). Recommendations were made to undertake actions to improve the understanding of the incident sequence and of the consequence this had on the fuel. It was considered that the Paks-2 event may constitute a useful case for a comparative exercise on safety codes, in particular for models devised to predict fuel damage and potential releases under abnormal cooling conditions and the analyses of the Paks-2 event may provide information which is relevant for in-reactor and spent fuel storage safety evaluations. The OECD-IAEA Paks Fuel Project was established in 2005 as a joint project between the IAEA and the OECD/NEA. The IAEA

  6. Fuel assembly leakage, unit 4, cycle 22, Paks NPP

    International Nuclear Information System (INIS)

    Szecsenyi, C.; Burjan, T.; Torma, B.; Bona, G.

    2009-01-01

    At the beginning of Cycle 22, Unit 4, Paks NPP the Iodine isotopes activity concentrations raised irregularly in the water of the primary circuit. Analysis supposed that from 1 to 10 fuel rods in one or more newly loaded follower assemblies had lost their integrity. Due to the fact it was not necessary to shut down the reactor, but at the end of the cycle sipping tests were performed for the entire core to find out the facts using a telescope sipping device supplied by H and B Co., Germany. This paper describes the circumstances of the emergence of the problem, the operational inspection and limitation rules in Paks NPP, the theoretical analysis to estimate the scope and kind of the problem, the sipping device and the measurement/evaluation methods applied for the practical tests, fulfilment the tests, the results and their evaluation and the conclusions regarding the event. (Authors)

  7. Fuel assembly leakage, Unit 4, Cycle 22, Paks NPP

    International Nuclear Information System (INIS)

    Szecsenyi, Z.; Burjan, T.; Torma, B.; Bona, G.

    2009-01-01

    At the beginning of Cycle 22, Unit 4, Paks NPP the Iodine isotopes activity concentrations raised irregularly in the water of the primary circuit. Analysis supposed that from 1 to 10 fuel rods in one or more newly loaded follower assemblies had lost their integrity. Due to the fact it was not necessary to shut down the reactor, but at the end of the cycle sipping tests were performed for the entire core to find out the facts using a telescope sipping device supplied by H and B Co., Germany. This paper describes the circumstances of the emergence of the problem, the operational inspection and limitation rules in the Paks NPP, the theoretical analysis to estimate the scope and kind of the problem, the sipping device and the measurement/evaluation methods applied for the practical tests, fulfilment the tests, the results and their evaluation and the conclusions regarding the event. (authors)

  8. Calculation uncertainty of distribution-like parameters in NPP of PAKS

    International Nuclear Information System (INIS)

    Szecsenyi, Zsolt; Korpas, Layos

    2000-01-01

    In the reactor-physical point of view there were two important events in the Nuclear Power Plant of PAKS in this year. The Russian type profiled assemblies were loaded into the PAKS Unit 3, and new limitation system was introduced on the same Unit. It was required to solve a lot of problems because of these both events. One of these problems was the determination of uncertainty of quantities of the new limitation considering the fabrication uncertainties for the profiled assembly. The importance of determination of uncertainty is to guarantee on 99.9% level the avoidance of fuel failure. In this paper the principles of determination of calculation accuracy, applied methods and obtained results are presented in case of distribution-like parameters. A few elements of the method have been presented on earlier symposiums, so in this paper the whole method is just outlined. For example the GPT method was presented in the following paper: Uncertainty analysis of pin wise power distribution of WWER-440 assembly considering fabrication uncertainties. Finally in the summary of this paper additional intrinsic opportunities in the method are presented. (Authors)

  9. Lifetime-management and lifetime-extension at PAKS nuclear power plant

    International Nuclear Information System (INIS)

    Katona, Tamas; Ratkai, Sandor; Janosi, Agnes Biro

    2002-01-01

    Paks Nuclear Power Plant provides 38-40% of domestic generation at lowest price. It has an important energy-policy role in Hungary. NPP Paks shall be a decisive and perspectively permanent element of the domestic electricity generation during the next two decades, which shall be ensured by plant safe operation, the lifetime extension and power uprating. Paks Nuclear Power Plant investigated the nuclear power plant's lifetime extension possibilities and alternatives, as well as technical and business feasibility of such alternatives. The feasibility study is based on the evaluation of a representative set of systems, structures and components, operational, test, in-service inspection and maintenance practice, experience and findings of the Periodic Safety Review. The most important results of this study showing the feasibility of 20 years lifetime extension is summarised in the paper. It was found that there are no technical or safety issues or limits, which may inhibit the operation of the Nuclear Power Plant Paks up to 50 years. In case of most systems and equipment the recent monitoring, maintenance and regular reconstruction practice of the NPP Paks allows the lifetime extension without outstanding cost. Replacement or reconstruction of a few equipment and systems requires significant investment costs. Material of reactor vessels of VVER/213 incorporated at Paks, compared to vessels of the similar units, is less sensitive to the embrittlement. At units 3-4 reactor vessels do not require any measure, consequently, any additional cost, even in case of a lifetime of 50 years. At unit 2 to extend the lifetime of the reactor vessel, only heating-up of emergency core cooling tanks is needed in order to decrease thermal stress levels caused by pressure thermal shock (PST) transients. For this purpose cost-effective technical solutions are available. At unit 1, beside the heating-up of the emergency core cooling tanks annealing of the welded joint No. 5/6 close to the

  10. Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

    International Nuclear Information System (INIS)

    Hozer, Zoltan; Szabo, Emese; Pinter, Tamas; Varju, Ilona Baracska; Bujtas, Tibor; Farkas, Gabor; Vajda, Nora

    2009-01-01

    During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.

  11. Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Zoltan, E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest 114, P.O. Box 49 (Hungary); Szabo, Emese [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest 114, P.O. Box 49 (Hungary); Pinter, Tamas; Varju, Ilona Baracska; Bujtas, Tibor; Farkas, Gabor [Nuclear Power Plant Paks, H-7031 Paks, P.O. Box 71 (Hungary); Vajda, Nora [Institute of Nuclear Techniques, Budapest University of Technology and Economics, H-1521 Budapest, Muegyetem rakpart 9 (Hungary)

    2009-07-01

    During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.

  12. The p21-activated kinase (PAK family member PakD is required for chemorepulsion and proliferation inhibition by autocrine signals in Dictyostelium discoideum.

    Directory of Open Access Journals (Sweden)

    Jonathan E Phillips

    Full Text Available In Dictyostelium discoideum, the secreted proteins AprA and CfaD function as reporters of cell density and regulate cell number by inhibiting proliferation at high cell densities. AprA also functions to disperse groups of cells at high density by acting as a chemorepellent. However, the signal transduction pathways associated with AprA and CfaD are not clear, and little is known about how AprA affects the cytoskeleton to regulate cell movement. We found that the p21-activated kinase (PAK family member PakD is required for both the proliferation-inhibiting activity of AprA and CfaD and the chemorepellent activity of AprA. Similar to cells lacking AprA or CfaD, cells lacking PakD proliferate to a higher cell density than wild-type cells. Recombinant AprA and CfaD inhibit the proliferation of wild-type cells but not cells lacking PakD. Like AprA and CfaD, PakD affects proliferation but does not significantly affect growth (the accumulation of mass on a per-nucleus basis. In contrast to wild-type cells, cells lacking PakD are not repelled from a source of AprA, and colonies of cells lacking PakD expand at a slower rate than wild-type cells, indicating that PakD is required for AprA-mediated chemorepulsion. A PakD-GFP fusion protein localizes to an intracellular punctum that is not the nucleus or centrosome, and PakD-GFP is also occasionally observed at the rear cortex of moving cells. Vegetative cells lacking PakD show excessive actin-based filopodia-like structures, suggesting that PakD affects actin dynamics, consistent with previously characterized roles of PAK proteins in actin regulation. Together, our results implicate PakD in AprA/CfaD signaling and show that a PAK protein is required for proper chemorepulsive cell movement in Dictyostelium.

  13. The p21-activated kinase (PAK) family member PakD is required for chemorepulsion and proliferation inhibition by autocrine signals in Dictyostelium discoideum.

    Science.gov (United States)

    Phillips, Jonathan E; Gomer, Richard H

    2014-01-01

    In Dictyostelium discoideum, the secreted proteins AprA and CfaD function as reporters of cell density and regulate cell number by inhibiting proliferation at high cell densities. AprA also functions to disperse groups of cells at high density by acting as a chemorepellent. However, the signal transduction pathways associated with AprA and CfaD are not clear, and little is known about how AprA affects the cytoskeleton to regulate cell movement. We found that the p21-activated kinase (PAK) family member PakD is required for both the proliferation-inhibiting activity of AprA and CfaD and the chemorepellent activity of AprA. Similar to cells lacking AprA or CfaD, cells lacking PakD proliferate to a higher cell density than wild-type cells. Recombinant AprA and CfaD inhibit the proliferation of wild-type cells but not cells lacking PakD. Like AprA and CfaD, PakD affects proliferation but does not significantly affect growth (the accumulation of mass) on a per-nucleus basis. In contrast to wild-type cells, cells lacking PakD are not repelled from a source of AprA, and colonies of cells lacking PakD expand at a slower rate than wild-type cells, indicating that PakD is required for AprA-mediated chemorepulsion. A PakD-GFP fusion protein localizes to an intracellular punctum that is not the nucleus or centrosome, and PakD-GFP is also occasionally observed at the rear cortex of moving cells. Vegetative cells lacking PakD show excessive actin-based filopodia-like structures, suggesting that PakD affects actin dynamics, consistent with previously characterized roles of PAK proteins in actin regulation. Together, our results implicate PakD in AprA/CfaD signaling and show that a PAK protein is required for proper chemorepulsive cell movement in Dictyostelium.

  14. P21-activated kinase 2 (PAK2) regulates glucose uptake and insulin sensitivity in neuronal cells.

    Science.gov (United States)

    Varshney, Pallavi; Dey, Chinmoy Sankar

    2016-07-05

    P21-activated kinases (PAKs) are recently reported as important players of insulin signaling and glucose homeostasis in tissues like muscle, pancreas and liver. However, their role in neuronal insulin signaling is still unknown. Present study reports the involvement of PAK2 in neuronal insulin signaling, glucose uptake and insulin resistance. Irrespective of insulin sensitivity, insulin stimulation decreased PAK2 activity. PAK2 downregulation displayed marked enhancement of GLUT4 translocation with increase in glucose uptake whereas PAK2 over-expression showed its reduction. Treatment with Akti-1/2 and wortmannin suggested that Akt and PI3K are mediators of insulin effect on PAK2 and glucose uptake. Rac1 inhibition demonstrated decreased PAK2 activity while inhibition of PP2A resulted in increased PAK2 activity, with corresponding changes in glucose uptake. Taken together, present study demonstrates an inhibitory role of insulin signaling (via PI3K-Akt) and PP2A on PAK2 activity and establishes PAK2 as a Rac1-dependent negative regulator of neuronal glucose uptake and insulin sensitivity. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

  15. 1α,25(OH2D3 Induces Actin Depolymerization in Endometrial Carcinoma Cells by Targeting RAC1 and PAK1

    Directory of Open Access Journals (Sweden)

    Ni Zeng

    2016-12-01

    Full Text Available Background: Cell proliferation and motility require actin reorganization, which is under control of various signalling pathways including ras-related C3 botulinum toxin substrate 1 (RAC1, p21 protein-activated kinase 1 (PAK1 and actin related protein 2 (ARP2. Tumour cell proliferation is modified by 1α,25-Dihydroxy-Vitamin D3 (1α,25(OH2D3, a steroid hormone predominantly known for its role in calcium and phosphorus metabolism. The present study explored whether 1α,25(OH2D3 modifies actin cytoskeleton in Ishikawa cells, a well differentiated endometrial carcinoma cell line. Methods: To this end, actin cytoskeleton was visualized by confocal microscopy. Globular over filamentous actin ratio was determined utilizing Western blotting and flow cytometry, transcript levels by qRT-PCR and protein abundance by immunoblotting. Results: A 24 hour treatment with 1α,25(OH2D3 (100 nM significantly decreased RAC1 and PAK1 transcript levels and activity, decreased ARP2 protein levels and depolymerized actin. The effect of 1α,25(OH2D3 on actin polymerization was mimicked by pharmacological inhibition of RAC1 and PAK1. Conclusions: 1α,25(OH2D3 leads to disruption of RAC1 and PAK1 activity with subsequent actin depolymerization of endometrial carcinoma cells.

  16. AGNES - safety reassessment of Paks NPP

    International Nuclear Information System (INIS)

    Gado, J.

    1995-01-01

    The main goal of the AGNES (Advanced General and New Evaluation of Safety) project for the reassessment of the safety of Paks Nuclear Power Plant, Hungary, was to improve the safety culture of the technology at Paks. A report was prepared on the reassessment of the Paks NPP safety. The analysis was divided into four groups: systems analysis, analysis of design basis accidents, severe accident analysis, and level 1 probabilistic safety analysis. Proposed safety enhancement measures are discussed. (N.T.)

  17. Full scale dynamic testing of Paks nuclear power plant structures

    International Nuclear Information System (INIS)

    Da Rin, E.M.

    1995-01-01

    This report refers to the full-scale dynamic structural testing activities that have been performed in December 1994 at the Paks (H) Nuclear Power Plant, within the framework of: the IAEA Coordinated research Programme 'Benchmark Study for the Seismic Analysis and Testing of WWER-type Nuclear Power Plants, and the nuclear research activities of ENEL-WR/YDN, the Italian National Electricity Board in Rome. The specific objective of the conducted investigation was to obtain valid data on the dynamic behaviour of the plant's major constructions, under normal operating conditions, for enabling an assessment of their actual seismic safety to be made. As described in more detail hereafter, the Paks NPP site has been subjected to low level earthquake like ground shaking, through appropriately devised underground explosions, and the dynamic response of the plant's 1 st reactor unit important structures was appropriately measured and digitally recorded. In-situ free field response was measured concurrently and, moreover, site-specific geophysical and seismological data were simultaneously acquired too. The above-said experimental data is to provide basic information on the geophysical and seismological characteristics of the Paks NPP site, together with useful reference information on the true dynamic characteristics of its main structures and give some indications on the actual dynamic soil-structure interaction effects for the case of low level excitation

  18. Rac1-PAK2 pathway is essential for zebrafish heart regeneration

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Xiangwen [State Key Laboratory of Genetic Engineering, Department of Genetics, School of Life Sciences, Fudan University, Shanghai 201203 (China); He, Quanze [Center for Reproduction and Genetics, Suzhou Municipal Hospital, Jiangsu 215002 (China); Li, Guobao; Ma, Jinmin [State Key Laboratory of Genetic Engineering, Department of Genetics, School of Life Sciences, Fudan University, Shanghai 201203 (China); Zhong, Tao P., E-mail: taozhongfudan@yahoo.com [State Key Laboratory of Genetic Engineering, Department of Genetics, School of Life Sciences, Fudan University, Shanghai 201203 (China); Department of Medicine, Vanderbilt University School of Medicine, TN 37232 (United States)

    2016-04-15

    P-21 activated kinases, or PAKs, are serine–threonine kinases that play important roles in diverse heart functions include heart development, cardiovascular development and function in a range of models; however, the mechanisms by which PAKs mediate heart regeneration are unknown. Here, we demonstrate that PAK2 and PAK4 expression is induced in cardiomyocytes and vessels, respectively, following zebrafish heart injury. Inhibition of PAK2 and PAK4 using a specific small molecule inhibitor impedes cardiomyocyte proliferation/dedifferentiation and cardiovascular regeneration, respectively. Cdc42 is specifically expressed in the ventricle and may function upstream of PAK2 but not PAK4 under normal conditions and that cardiomyocyte proliferentation during heart regeneration relies on Rac1-mediated activation of Pak2. Our results indicate that PAKs play a key role in heart regeneration.

  19. Rac1-PAK2 pathway is essential for zebrafish heart regeneration

    International Nuclear Information System (INIS)

    Peng, Xiangwen; He, Quanze; Li, Guobao; Ma, Jinmin; Zhong, Tao P.

    2016-01-01

    P-21 activated kinases, or PAKs, are serine–threonine kinases that play important roles in diverse heart functions include heart development, cardiovascular development and function in a range of models; however, the mechanisms by which PAKs mediate heart regeneration are unknown. Here, we demonstrate that PAK2 and PAK4 expression is induced in cardiomyocytes and vessels, respectively, following zebrafish heart injury. Inhibition of PAK2 and PAK4 using a specific small molecule inhibitor impedes cardiomyocyte proliferation/dedifferentiation and cardiovascular regeneration, respectively. Cdc42 is specifically expressed in the ventricle and may function upstream of PAK2 but not PAK4 under normal conditions and that cardiomyocyte proliferentation during heart regeneration relies on Rac1-mediated activation of Pak2. Our results indicate that PAKs play a key role in heart regeneration.

  20. Diagnostics of the boiling state of coolant based on neutron fluctuation at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Por, G.; Gloeckler, O.; Izsak, E.; Valko, J.

    1985-09-01

    A short summary of theory and early experiments on the effect of propagating perturbation on neutron fluctuations in nuclear reactors is given. Boiling noise was examined in the Rheisenberg reactor of 70 MWe. Comparing the results of measurements with those carried out in the Paks nuclear power plant it seems possible that a small subcooled boiling took place during the 2nd fuel cycle. (author)

  1. Group I Paks Promote Skeletal Myoblast Differentiation In Vivo and In Vitro

    DEFF Research Database (Denmark)

    Joseph, Giselle A; Lu, Min; Radu, Maria

    2017-01-01

    fusion in Drosophila We report that both Pak1 and Pak2 are activated during mammalian myoblast differentiation. One pathway of activation is initiated by N-cadherin ligation and involves the cadherin coreceptor Cdo with its downstream effector, Cdc42. Individual genetic deletion of Pak1 and Pak2 in mice....... Furthermore, primary myoblasts lacking Pak1 and Pak2 display delayed expression of myogenic differentiation markers and myotube formation. These results identify Pak1 and Pak2 as redundant regulators of myoblast differentiation in vitro and in vivo and as components of the promyogenic Ncad/Cdo/Cdc42 signaling...

  2. Seismic analyses of Paks RB. Progress report 1993-1994

    Energy Technology Data Exchange (ETDEWEB)

    David, M [David Consulting, Engineering and Design Office (Czech Republic)

    1995-07-01

    The dynamic analysis presented in this report refers to the seismic analysis of the main building of Paks NPP. The following tasks which have been completed are described: design of 3-dimensional model of the main building; calculation of frequencies and modes of free vibrations; determination of modal masses for all modes of vibrations; floor response spectra as response to seismic excitation assumed for the Paks site; relative response of seismic acceleration at the top of the condensing tower.

  3. A modernized and versatile startup reactivity measuring system installed at NPP Paks and its application for subcritical systems

    International Nuclear Information System (INIS)

    Czibok, T.; Dezso, Z.; Horvath, Cs.; Lipcsei, S.; Vegh, J.; Pos, I.

    2006-01-01

    In 2004 the Hungarian Paks NPP completed a project for upgrading the reactivity measuring system applied during reactor startup experiments. Almost all components of the previous system were replaced, only ex-core ionisation chambers remained unaltered. New hardware and software components were introduced for neutron flux signal handling, for data acquisition, as well as for measurement evaluation and data presentation. High-precision picoamper meters were installed at each reactor unit, current signals are handled by a portable signal processing unit. The system applies an accurate on-line reactivity calculation algorithm based on the point-kinetic model with six delayed neutron groups. Detailed off-line evaluation and analysis of startup measurements can be performed on the portable unit, as well. The paper describes the architecture, data acquisition modules, services and man-machine interface of the new system. Functions and results are illustrated with measured data recorded during a startup of Unit 3. In 2003 and 2004 the RMR was installed and tested at all Paks NPP units successfully and now it is in regular use during unit startups. The second part of the paper illustrates an extension of the new system to perform reactivity measurements using the well-known Rossi-α and Feynman-α statistical methods. The modified system was needed to estimate the reactivity of a subcritical system formed by damaged fuel assemblies stored at the fuel service pit of Paks Unit 2. Theoretical background of the applied algorithms is outlined, then results of validation tests and on site measurements are treated. The measurements have shown that the subcriticality of the damaged fuel was sufficiently deep if the high boron concentration in the fuel service pit was maintained

  4. ASSET experience at Paks NPP

    International Nuclear Information System (INIS)

    Szabo, I.

    1997-01-01

    At Paks NPP special attention has been paid to international reviews since the very beginning of operation. Several international teams visited Paks in order to provide independent assessment of plant performance, conditions and safety. Paks NPP Management has the further intention to invite international reviews regularly (yearly) in the future as well. The experience gained during these reviews helped to establish a unified process of preparation for the reviews, performing them and handling the results. The Safety Department is in charge of organization of the whole process. All these reviews have their specific features and they are focused on different areas. The ASSET reviews provides the assessment of plant performance and safety through the analysis of safety significant events, which have occurred at the nuclear power plant. This approach makes this review specific and different from the other ones

  5. ASSET experience at Paks NPP

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, I [Operational Safety Dept., Paks NPP, Paks (Hungary)

    1997-10-01

    At Paks NPP special attention has been paid to international reviews since the very beginning of operation. Several international teams visited Paks in order to provide independent assessment of plant performance, conditions and safety. Paks NPP Management has the further intention to invite international reviews regularly (yearly) in the future as well. The experience gained during these reviews helped to establish a unified process of preparation for the reviews, performing them and handling the results. The Safety Department is in charge of organization of the whole process. All these reviews have their specific features and they are focused on different areas. The ASSET reviews provides the assessment of plant performance and safety through the analysis of safety significant events, which have occurred at the nuclear power plant. This approach makes this review specific and different from the other ones.

  6. Development of a Hydronic Rooftop Unit-HyPak-MA

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Eric; Berman, Mark

    2009-11-14

    The majority of U.S. commercial floor space is cooled by rooftop HVAC units (RTUs). RTU popularity derives chiefly from their low initial cost and relative ease of service access without disturbing building occupants. Unfortunately, current RTUs are inherently inefficient due to a combination of characteristics that unnecessarily increase cooling loads and energy use. 36% percent of annual U.S. energy, and two-thirds of electricity, is consumed in and by buildings. Commercial buildings consume approximately 4.2 quads of energy each year at a cost of $230 billion per year, with HVAC equipment consuming 1.2 quads of electricity. More than half of all U.S. commercial floor space is cooled by packaged HVAC units, most of which are rooftop units (RTUs). Inefficient RTUs create an estimated 3.5% of U.S. CO{sub 2} emissions, thus contributing significantly to global warming5. Also, RTUs often fail to maintain adequate ventilation air and air filtration, reducing indoor air quality. This is the second HyPak project to be supported by DOE through NETL. The prior project, referred to as HyPak-1 in this report, had two rounds of prototype fabrication and testing as well as computer modeling and market research. The HyPak-1 prototypes demonstrated the high performance capabilities of the HyPak concept, but made it clear that further development was required to reduce heat exchanger cost and improve system reliability before HyPak commercialization can commence. The HyPak-1 prototypes were limited to about 25% ventilation air fraction, limiting performance and marketability. The current project is intended to develop a 'mixed-air' product that is capable of full 0-100% modulation in ventilation air fraction, hence it was referred to as HyPak-MA in the proposal. (For simplicity, the -MA has been dropped when referencing the current project.) The objective of the HyPak Project is to design, develop and test a hydronic RTU that provides a quantum improvement over

  7. Cerita Humor Pak Andir

    Directory of Open Access Journals (Sweden)

    Rohim Rohim

    2014-06-01

    Full Text Available This study attempts to describe the meaning of comic tale "Pak Andir" with the perspective of hermeneutics. This study is focused on exploring the main character with the theory of functional models and aktan, developed by Greimas. The source of data is the story of "Pak Andir" from the community of South Bengkulu. From the analysis, it is concluded that the behavior of the husband as the central character has made the wife a victim. The husband’s arrogance in strictly practicing the patriarchal tradition makes the wife have no courage to be herself. The wife’s claim at the end of the story is a positive thing, but it's too late. As a form of appreciation of literary work, the meaning of these stories need to be disseminated to the public, especially the residents in Bengkulu, that the husband and wife’s attitudes ares incorrect and need to be avoided. This study attempts to describe the meaning of comic tale "Pak Andir" with the perspective of hermeneutics. This study is focused on exploring the main character with the theory of functional models and aktan, developed by Greimas. The source of data is the story of "Pak Andir" from the community of South Bengkulu. From the analysis, it is concluded that the behavior of the husband as the central character has made the wife a victim. The husband’s arrogance in strictly practicing the patriarchal tradition makes the wife have no courage to be herself. The wife’s claim at the end of the story is a positive thing, but it's too late. As a form of appreciation of literary work, the meaning of these stories need to be disseminated to the public, especially the residents in Bengkulu, that the husband and wife’s attitudes ares incorrect and need to be avoided Key Words: comic tale; aktan model; functional model; hermeneutic Abstrak: Penelitian ini berusaha mendeskripsikan makna cerita humor “Pak Andir” dengan perspektif hermeneutika. Kajian ini difokuskan untuk mengeksplorasi tokoh utama

  8. Proposal of In-vessel corium retention concept for Paks NPP

    International Nuclear Information System (INIS)

    Elter, J.; Toth, E.; Matejovic, P.

    2011-01-01

    The in-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) seems to be a promising severe accident management strategy not only for new generation of advanced PWRs, but also for VVER-440/V213 reactors, which were designed several years ago. The basic idea of in-vessel retention of corium is to prevent RPV failure by flooding the reactor cavity so that the reactor pressure vessel is submerged in water up to its support structures, and thus the decay heat can be transferred from the corium pool through the vessel wall and into the water surrounding the vessel. An IVR concept with simple ECVR loop based only on minor modifications of existing plant technology was proposed for the Paks Nuclear Power Plant. 2 severe accident (LB and SB LOCA) without availability of HP and LP safety injection in power upgrade (108%) conditions were simulated using the ASTEC code. The analyses show that the proposed solution is effective in preserving RPV integrity in the case of severe accident. Possible uncertainties in code predictions are covered by the applied conservative assumptions

  9. Safety improvement of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Vamos, G.

    1999-01-01

    Safety upgrading completed in the early nineties at the Paks NPP include: replacement of steam generator safety valves and control valves; reliability improvement of the electrical supply system; modification of protection logic; enhancement of the fire protection; construction of full scope Training Simulator. Design safety upgrading measures achieved in recent years were concerned with: relocation of steam generator emergency feed-water supply; emergency gas removal from the primary coolant system; hydrogen management in the containment; protection against sumps; preventing of emergency core cooling system tanks from refilling. Increasing seismic resistance, containment assessment, refurbishment of reactor protection system, improving reliability of emergency electrical supply, analysis of internal hazards are now being implemented. Safety upgrading measures which are being prepared include: bleed and feed procedures; reactor over-pressurisation protection in cold state; treatment of steam generator primary to secondary leak accidents. Operational safety improvements are dealing with safety culture, training measures and facilities; symptom based emergency operating procedures; in-service inspection; fire protection. The significance of international cooperation is emphasised in view of achieving nuclear safety standards recognised in EU

  10. MODELACIÓN DEL PROCESO DE RECUPERACIÓN PARCIAL DE ENVASES DE TETRA PAK MODELAÇÃO DO PROCESSO DE RECUPERAÇÃO PARCIAL DE EMBALAGENS TETRA PAK MODELING THE PARTIAL RECOVERY PROCESS OF TETRA PAK PACKAGES

    Directory of Open Access Journals (Sweden)

    JORGE MARIO OBANDO

    2009-07-01

    Full Text Available En el presente artículo se estudian el patrón de consumo, las expectativas y satisfacción de la población de Medellín con los envases de Tetra Pak, el tratamiento que se les da cuando son descartados y la conducta que se seguiría, conociendo que el Tetra Pak es reciclable; luego, se modela el procedimiento para recuperar parcialmente esos desechos de Tetra Pak y comercializarlos como bienes intermedios, aptos para ser reintegrados dentro de diversos procesos productivos. Se estudian varias políticas de operación mediante un modelo de simulación de eventos discretos construido en Extend.No presente artigo se estudam o patrão de consumo, as expectativas e satisfação da população de M edellín com as embalagens Tetra Pak, o tratamento que se lhes dá quando são descartadas e a conduta que se seguiria, conhecendo que o Tetra Pak é reciclável; depois, se modela o procedimento para recuperar parcialmente esses resíduos de Tetra Pak e comercializá-los como bens intermédios, aptos para ser reintegrados dentro de diversos processos produtivos. Estudam-se várias políticas de operação mediante um modelo de simulação de eventos discretos construído em Extend.In this paper we study the consumption pattern, expectations and satisfaction of the population of Medellin with Tetra Pak packages, the treatment given to them now when they are discarded and the conduct to be followed, knowing that the Tetra Pak is recyclable, then the process to recover partially debris from Tetra Pak is modeled and marketed as intermediate goods, eligible to be reinstated within various production processes. Several operating policies through a discrete event simulation built in Extend are studied.

  11. Research and higher education background of the Paks Nuclear Power Plant, Hungary. Past and present

    International Nuclear Information System (INIS)

    Csom, Gy.

    2002-01-01

    The connection of the Paks Nuclear Power Plant, Hungary, with research and development as well as with higher education is discussed. The main research areas include reactor physics, thermohydraulics, radiochemistry and radiochemical analysis, electronics and nuclear instruments, computers, materials science. The evolution of relations with higher education in Hungary and the PNPP is presented, before and after the installation of the various units. (R.P.)

  12. miR-129 suppresses tumor cell growth and invasion by targeting PAK5 in hepatocellular carcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Jian [Department II of Interventional Radiology, Eastern Hepatobiliary Surgery Hospital, Second Military Medical University, Shanghai 200438 (China); Qu, Shuping [Department II of Special Medical Care, Eastern Hepatobiliary Surgery Hospital, Second Military Medical University, Shanghai 200438 (China); Li, Xiaowei; Zhong, Jiaming; Chen, Xiaoxia [Department II of Interventional Radiology, Eastern Hepatobiliary Surgery Hospital, Second Military Medical University, Shanghai 200438 (China); Qu, Zengqiang, E-mail: drquzengqiang@163.com [Department II of Interventional Radiology, Eastern Hepatobiliary Surgery Hospital, Second Military Medical University, Shanghai 200438 (China); Wu, Dong, E-mail: wudongstc@126.com [Department II of Special Medical Care, Eastern Hepatobiliary Surgery Hospital, Second Military Medical University, Shanghai 200438 (China)

    2015-08-14

    Emerging evidence suggests that microRNAs (miRNAs) play important roles in regulating HCC development and progression; however, the mechanisms by which their specific functions and mechanisms remained to be further explored. miR-129 has been reported in gastric cancers, lung cancer and colon cancer. In this study, we disclosed a new tumor suppresser function of miR-129 in HCC. We also found the downregulation of miR-129 occurred in nearly 3/4 of the tumors examined (56/76) compared with adjacent nontumorous tissues, which was more importantly, correlated to the advanced stage and vascular invasion. We then demonstrated that miR-129 overexpression attenuated HCC cells proliferation and invasion, inducing apoptosis in vitro. Moreover, we used miR-129 antagonist and found that anti-miR-129 promoted HCC cells malignant phenotypes. Mechanistically, our further investigations revealed that miR-129 suppressed cell proliferation and invasion by targeting the 3’-untranslated region of PAK5, as well as miR-129 silencing up-regulated PAK5 expression. Moreover, miR-129 expression was inversely correlated with PAK5 expression in 76 cases of HCC samples. RNA interference of PAK5 attenuated anti-miR-129 mediated cell proliferation and invasion in HCC cells. Taken together, these results demonstrated that miR-129 suppressed tumorigenesis and progression by directly targeting PAK5, defining miR-129 as a potential treatment target for HCC. - Highlights: • Decreased of miR-129 is found in HCC and associated with advanced stage and metastasis. • miR-129 suppresses proliferation and invasion of HCC cells. • miR-129 directly targets the 3′ UTR of PAK5 and diminishes PAK5 expression. • PAK5 is involved in miR-129 mediated suppression functions.

  13. GIT1/βPIX signaling proteins and PAK1 kinase regulate microtubule nucleation.

    Science.gov (United States)

    Černohorská, Markéta; Sulimenko, Vadym; Hájková, Zuzana; Sulimenko, Tetyana; Sládková, Vladimíra; Vinopal, Stanislav; Dráberová, Eduarda; Dráber, Pavel

    2016-06-01

    Microtubule nucleation from γ-tubulin complexes, located at the centrosome, is an essential step in the formation of the microtubule cytoskeleton. However, the signaling mechanisms that regulate microtubule nucleation in interphase cells are largely unknown. In this study, we report that γ-tubulin is in complexes containing G protein-coupled receptor kinase-interacting protein 1 (GIT1), p21-activated kinase interacting exchange factor (βPIX), and p21 protein (Cdc42/Rac)-activated kinase 1 (PAK1) in various cell lines. Immunofluorescence microscopy revealed association of GIT1, βPIX and activated PAK1 with centrosomes. Microtubule regrowth experiments showed that depletion of βPIX stimulated microtubule nucleation, while depletion of GIT1 or PAK1 resulted in decreased nucleation in the interphase cells. These data were confirmed for GIT1 and βPIX by phenotypic rescue experiments, and counting of new microtubules emanating from centrosomes during the microtubule regrowth. The importance of PAK1 for microtubule nucleation was corroborated by the inhibition of its kinase activity with IPA-3 inhibitor. GIT1 with PAK1 thus represent positive regulators, and βPIX is a negative regulator of microtubule nucleation from the interphase centrosomes. The regulatory roles of GIT1, βPIX and PAK1 in microtubule nucleation correlated with recruitment of γ-tubulin to the centrosome. Furthermore, in vitro kinase assays showed that GIT1 and βPIX, but not γ-tubulin, serve as substrates for PAK1. Finally, direct interaction of γ-tubulin with the C-terminal domain of βPIX and the N-terminal domain of GIT1, which targets this protein to the centrosome, was determined by pull-down experiments. We propose that GIT1/βPIX signaling proteins with PAK1 kinase represent a novel regulatory mechanism of microtubule nucleation in interphase cells. Copyright © 2016 Elsevier B.V. All rights reserved.

  14. Continuous analysis of radioiodine isotopes in the primary coolant of NPP Paks, Hungary

    International Nuclear Information System (INIS)

    Erdoes, E.; Soos, J.; Vincze, A.; Zsille, O.; Gujgiczer, A.; Solymosi, J.; Pinter, T.

    1998-01-01

    The radioiodine analyser has been installed at the Paks-3 reactor unit. The analyser is based on an efficient and simple method of radioiodine separation: the iodine compound is converted to elementary iodine quantitatively by oxidation with potassium iodate in acid medium. Owing to its volatility, iodine is evaporated quantitatively from the primary coolant (desorption) using air flow. The air is bubbled through a solution of a reducer, and iodine is absorbed in a form which is ready for measurement. A simple NaI(Tl) detector is used for the measurement of gamma spectra. The system is controlled and data are processed by a computer. The analyser displays activity concentration data of the five iodine isotopes periodically every 15 minutes. (M.D.)

  15. Safety upgrading of the PAKS Nuclear Plant

    International Nuclear Information System (INIS)

    Vamos, G.; Vigassy, J.

    1993-01-01

    In the last several years the net electricity from the Paks NPP represents almost half of the Hungarian total. The 4 units of Paks belong to the latest generation of the VVER-440 units, the small-sized Russian designed PWRs. Reviewing the main design features of them, the safety merits and safety concerns are summarized. Due to the conservative design and the extensive operating experience the safety merits appear to be more significant than generally believed. The VVER-440 type has two models, the 230 and 213, which have a large number of distinctive safety features. These are highlighted in the section comparisons. A quality assurance program was initiated in Paks very early. A long-term safety upgrading program was also initiated, originating from vendor recommendations, regulatory decisions, in-house operating experience and safety concerns, and independent reviews. The main areas and some examples of the measures are described. This program, like all other activities related to nuclear safety, has been under regulatory control. The specific features of the Hungarian regulatory system are described. For advanced, general and new evaluation of the safety of the units in Paks in accordance with the internationally recommended criteria of the 90's, the project AGNES has been launched with international participation. The scope of this project is summarized. International efforts as the IAEA Regional Project on safety assessment of VVER-440/213 and VVER-440/230 units are underway. Since safety is not only a question of design, but it can be significantly influenced by operations and maintenance practices, the Paks NPP has invited LAEA's OSART and ASSET missions, WANO's Pilot Peer Review

  16. RMR. A new portable Reactivity Measuring System installed at NPP Paks

    International Nuclear Information System (INIS)

    Czibok, T.; Horvath, C.; Bara, P.; Dezsoe, Z.; Laz, J.; Vegh, J.; Pos, I.

    2003-01-01

    The Hungarian Paks NPP is conducting a two year project for upgrading the reactivity measuring system applied during reactor startup experiments. The NPP has decided to replace almost all components of the previous system, only ionisation chambers remain unaltered. Devices for measuring neutron flux by means of ionisation chambers, for data acquisition and for measurement evaluation were completely renewed: new hardware-software components were introduced. Autonomous, high-precision current measuring systems (picoampere meters) are applied at each reactor unit, the converted picoampere signals are handled by a portable processing unit. The portable unit - based on a notebook PC - handles measured signals by using a high-precision A/D converter card, the scan time is 0.10 sec. In addition to handling three ionisation chamber signals the portable unit collects control rod position measurements through a serial line. The portable unit is able to receive additional measured data (e.g. core inlet temperature and boron concentration) from the process computer via local area network. Archiving of all measured and calculated data is performed in a redundant manner: data are stored locally and in the process computer, as well. The new system applies an accurate on-line reactivity calculation algorithm based on the point-kinetic model with 6 delayed neutron groups. Input data (effective delayed neutron fraction and other delayed neutron parameters) to the on-line calculation are taken from the off-line core design calculation. Detailed evaluation and analysis of startup measurements can be performed also on the portable unit. The user interface of the system is tailored to support various startup measurement tasks effectively: measured and calculated data are displayed on trends and on dedicated pictures. A user-friendly trending and listing graphic tool facilitates visualisation of archived data. The paper describes the architecture, data acquisition modules, algorithms and

  17. PTS assessment - The basis of life time evaluation at NPP Paks

    International Nuclear Information System (INIS)

    Elter, J.; Oszwald, F.; Ratkay, S.; Fekete, T.; Gillemot, F.; Marothy, L.

    1997-01-01

    Plant specific PTS analysis at NPP Paks was performed in the frame of the AGNES (Advanced General New Evaluation of Safety) project. NPP Paks belongs to the second generation of the WWER-440/213 NPP-s. To verify the safety during transient events and predict the lifetime of the RPV-s several transient cases have been analyzed. The paper summarizes: The general scheme elaborated for the assessment; the safety philosophy used; the applied and available codes and methods; the ongoing and planned developments. (author). 8 refs, 3 figs, 1 tab

  18. PTS assessment - The basis of life time evaluation at NPP Paks

    Energy Technology Data Exchange (ETDEWEB)

    Elter, J; Oszwald, F; Ratkay, S [NPP Paks (Hungary); Fekete, T; Gillemot, F; Marothy, L [Atomic Energy Research Inst., Budapest (Hungary)

    1997-09-01

    Plant specific PTS analysis at NPP Paks was performed in the frame of the AGNES (Advanced General New Evaluation of Safety) project. NPP Paks belongs to the second generation of the WWER-440/213 NPP-s. To verify the safety during transient events and predict the lifetime of the RPV-s several transient cases have been analyzed. The paper summarizes: The general scheme elaborated for the assessment; the safety philosophy used; the applied and available codes and methods; the ongoing and planned developments. (author). 8 refs, 3 figs, 1 tab.

  19. Regulation of Stat5 by FAK and PAK1 in Oncogenic FLT3 and KIT driven Leukemogenesis

    Science.gov (United States)

    Chatterjee, Anindya; Ghosh, Joydeep; Ramdas, Baskar; Mali, Raghuveer Singh; Martin, Holly; Kobayashi, Michihiro; Vemula, Sasidhar; Canela, Victor H.; Waskow, Emily R.; Visconte, Valeria; Tiu, Ramon V.; Smith, Catherine C.; Shah, Neil; Bunting, Kevin D.; Boswell, H. Scott; Liu, Yan; Chan, Rebecca J.; Kapur, Reuben

    2015-01-01

    SUMMARY Oncogenic mutations of FLT3 and KIT receptors are associated with poor survival in patients with acute myeloid leukemia (AML) and myeloproliferative neoplasms (MPN) and currently available drugs are largely ineffective. Although Stat5 has been implicated in regulating several myeloid and lymphoid malignancies, how precisely Stat5 regulates leukemogenesis, including its nuclear translocation to induce gene transcription is poorly understood. In leukemic cells, we show constitutive activation of focal adhesion kinase (FAK), whose inhibition represses leukemogenesis. Downstream of FAK, activation of Rac1 is regulated by RacGEF Tiam1, whose inhibition prolongs the survival of leukemic mice. Inhibition of the Rac1 effector PAK1 prolongs the survival of leukemic mice in part by inhibiting the nuclear translocation of Stat5. These results reveal a leukemic pathway involving FAK/Tiam1/Rac1/PAK1 and demonstrate an essential role for these signaling molecules in regulating the nuclear translocation of Stat5 in leukemogenesis. PMID:25456130

  20. Regulation of Stat5 by FAK and PAK1 in Oncogenic FLT3- and KIT-Driven Leukemogenesis

    Directory of Open Access Journals (Sweden)

    Anindya Chatterjee

    2014-11-01

    Full Text Available Oncogenic mutations of FLT3 and KIT receptors are associated with poor survival in patients with acute myeloid leukemia (AML and myeloproliferative neoplasms (MPNs, and currently available drugs are largely ineffective. Although Stat5 has been implicated in regulating several myeloid and lymphoid malignancies, how precisely Stat5 regulates leukemogenesis, including its nuclear translocation to induce gene transcription, is poorly understood. In leukemic cells, we show constitutive activation of focal adhesion kinase (FAK whose inhibition represses leukemogenesis. Downstream of FAK, activation of Rac1 is regulated by RacGEF Tiam1, whose inhibition prolongs the survival of leukemic mice. Inhibition of the Rac1 effector PAK1 prolongs the survival of leukemic mice in part by inhibiting the nuclear translocation of Stat5. These results reveal a leukemic pathway involving FAK/Tiam1/Rac1/PAK1 and demonstrate an essential role for these signaling molecules in regulating the nuclear translocation of Stat5 in leukemogenesis.

  1. KRITIK SOSIAL DALAM KOMIK STRIP PAK BEI

    Directory of Open Access Journals (Sweden)

    Yudhi Novriansyah

    2016-08-01

    Full Text Available This research aimed to do interpret the marking which flange social criticism and know laboring ideology in story of Comic Strip Pak Bei. Research based on theory of structural semiotic according to Ferdinand De Saussure. Using analysis of Syntagmatic as first level of meaning to the text network and also picture, and analysis of Paradigmatic as second level of meaning or implicit meaning (connota-tion, myth, ideology Analysis done to six Comic choice edition of Strip Pak Bei period of November 2004 - Februari 2005 which tend to flange social criticism. At band of syntagmatic, result of research indicate that story theme lifted from social problems that happened in major society. The fact clear progressively when connected by Intertextual with information and texts which have preexisted. At band of Paradigmatic, social criticism tend to emerge dimly, is not transparent. Because of Comic Strip Pak Bei expand in the middle of Java cultural domination that developing myth of criticize as action menacing compatibility and orderliness of society. Story of Comic Strip Pak Bei also confirm dominant ideology in Java society culture, namely ideology of Patriarkhi and Feudalism which still go into effect until now. This prove ideology idea according to Louis Althusser which not again opposition between class, but have been owned and practiced by all social class.

  2. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4G. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  3. Introduction of the SAT based training programs at Paks NPP

    International Nuclear Information System (INIS)

    Kiss, I.

    1998-01-01

    An introduction of the SAT based training programs at Paks nuclear power plant is described in detail, including framework of project operation; project implementation; process of SAT applied at Paks NPP and the needs of its introduction

  4. Radiation protection measurements at Paks and its surroundings after the accident of the Chernobylsk nuclear power plant from 28 Apr 1986

    International Nuclear Information System (INIS)

    German, Endre; Kemenes, Laszlo; Rosa, Geza; Szabo, I.C.; Ormai, Peter; Ronaky, Jozsef; Divos, Ferenc; Varju, Bela; Horvath, Etelka.

    1986-08-01

    Experimental data on the contaminantion measured within a radius of 30 km from the Paks Nuclear Power Plant due to the accident of the Chernobylsk-4 reactor are given for the period between 28 Apr and 13 Jun 1986. Measurements on airborne and fallout activities, surface contamination of the ground, dose rates of γ radiation, activity concentration of the Danube, of milk, plant and food samples and the activity of human thyroid gland were carried out in the environmental control lab of the Paks NPP. According to the preliminary dose calculations the increment of the radiation exposure of the population is regarded to mount up to the average dose burden for half a year due to natural environmental radiation. (V.N.)

  5. Results of secondary side water regime modification in Nuclear Power Plant Paks

    International Nuclear Information System (INIS)

    Oesz, J.; Salamon, T.; Nagy, O.; Tilky, P.

    2001-01-01

    In order to extend the lifetime of Paks NPP, and for a possible power increase it is more and more evident that steam generators may be the limit. For the wear-out of the SG, it is decisive that at the end of the planned lifetime (after 25-30 reactor years) the number of plugged tubes should be as far as possible from the heat capacity limit. The modification of the secondary side water regime was started in 1997. It has been completed in the summer of year 2000, each of the four units has been operating using the new water regime. The results of this modification were evaluated on the basis of data obtained from six reactor years. The new water regime - after the overhaul check of the SG tubes - significantly decrease the number of tubes to plugged in the future. (R.P.)

  6. Regulation of Stat5 by FAK and PAK1 in Oncogenic FLT3- and KIT-Driven Leukemogenesis.

    Science.gov (United States)

    Chatterjee, Anindya; Ghosh, Joydeep; Ramdas, Baskar; Mali, Raghuveer Singh; Martin, Holly; Kobayashi, Michihiro; Vemula, Sasidhar; Canela, Victor H; Waskow, Emily R; Visconte, Valeria; Tiu, Ramon V; Smith, Catherine C; Shah, Neil; Bunting, Kevin D; Boswell, H Scott; Liu, Yan; Chan, Rebecca J; Kapur, Reuben

    2014-11-20

    Oncogenic mutations of FLT3 and KIT receptors are associated with poor survival in patients with acute myeloid leukemia (AML) and myeloproliferative neoplasms (MPNs), and currently available drugs are largely ineffective. Although Stat5 has been implicated in regulating several myeloid and lymphoid malignancies, how precisely Stat5 regulates leukemogenesis, including its nuclear translocation to induce gene transcription, is poorly understood. In leukemic cells, we show constitutive activation of focal adhesion kinase (FAK) whose inhibition represses leukemogenesis. Downstream of FAK, activation of Rac1 is regulated by RacGEF Tiam1, whose inhibition prolongs the survival of leukemic mice. Inhibition of the Rac1 effector PAK1 prolongs the survival of leukemic mice in part by inhibiting the nuclear translocation of Stat5. These results reveal a leukemic pathway involving FAK/Tiam1/Rac1/PAK1 and demonstrate an essential role for these signaling molecules in regulating the nuclear translocation of Stat5 in leukemogenesis. Copyright © 2014 The Authors. Published by Elsevier Inc. All rights reserved.

  7. PAK1 is a breast cancer oncogene that coordinately activates MAPK and MET signaling.

    Science.gov (United States)

    Shrestha, Y; Schafer, E J; Boehm, J S; Thomas, S R; He, F; Du, J; Wang, S; Barretina, J; Weir, B A; Zhao, J J; Polyak, K; Golub, T R; Beroukhim, R; Hahn, W C

    2012-07-19

    Activating mutations in the RAS family or BRAF frequently occur in many types of human cancers but are rarely detected in breast tumors. However, activation of the RAS-RAF-MEK-ERK MAPK pathway is commonly observed in human breast cancers, suggesting that other genetic alterations lead to activation of this signaling pathway. To identify breast cancer oncogenes that activate the MAPK pathway, we screened a library of human kinases for their ability to induce anchorage-independent growth in a derivative of immortalized human mammary epithelial cells (HMLE). We identified p21-activated kinase 1 (PAK1) as a kinase that permitted HMLE cells to form anchorage-independent colonies. PAK1 is amplified in several human cancer types, including 30--33% of breast tumor samples and cancer cell lines. The kinase activity of PAK1 is necessary for PAK1-induced transformation. Moreover, we show that PAK1 simultaneously activates MAPK and MET signaling; the latter via inhibition of merlin. Disruption of these activities inhibits PAK1-driven anchorage-independent growth. These observations establish PAK1 amplification as an alternative mechanism for MAPK activation in human breast cancer and credential PAK1 as a breast cancer oncogene that coordinately regulates multiple signaling pathways, the cooperation of which leads to malignant transformation.

  8. Comparative evaluation of anoxomat and conventional anaerobic GasPak jar systems for the isolation of anaerobic bacteria.

    Science.gov (United States)

    Shahin, May; Jamal, Wafaa; Verghese, Tina; Rotimi, V O

    2003-01-01

    To evaluate the performance of the Anoxomat, in comparison with the conventional anaerobic GasPak jar system, for the isolation of obligate anaerobes. Anoxomat, model WS800, and anaerobic GasPak jar system (Oxoid) were evaluated. Anoxomat system utilized a gas mixture of 80% N(2), 10% CO(2) and 10% H(2), while the GasPak used a gas mixture of 90% H(2) and 10% CO(2). An anaerobic indicator within the jars monitored anaerobiosis. A total of 227 obligate anaerobic bacteria comprising 116 stock strains, 5 ATCC reference strains and 106 fresh strains, representing different genera, were investigated for growth on anaerobic agar plates and scored for density, colony sizes, susceptibility zones of antibiotic inhibition and the speed of anaerobiosis (reducing the indicator). The results demonstrate that the growth of anaerobic bacteria is faster inside the Anoxomat jar than in the anaerobic GasPak jar system. Of the 227 strains tested, the colonies of 152 (67%) were larger (by size range of 0.2-2.4 mm) in the Anoxomat at 48 h than in the GasPak jar compared with only 21% (range 0.1-0.3 mm) that were larger in the GasPak than in the Anoxomat. The remaining 12% were equal in their sizes. There was no measurable difference in the colony sizes of the reference strains. The Porphyromonas asaccharolytica strains failed to grow within the GasPak system but grew inside the Anoxomat. With the Anoxomat, anaerobiosis was achieved about 35 min faster than in the GasPak system. The density of growth recorded for 177 (78%) strains was heavier in the Anoxomat than in the GasPak jar. The zones of inhibition of the antibiotics tested were not different in the two systems. The Anoxomat system provided superior growth, in terms of density and colony size, and achieved anaerobiosis more rapidly. Evidently, the Anoxomat method is more reliable and appears to support the growth of strict anaerobes better. Copyright 2003 S. Karger AG, Basel

  9. p21-Activated kinase (PAK regulates cytoskeletal reorganization and directional migration in human neutrophils.

    Directory of Open Access Journals (Sweden)

    Asako Itakura

    Full Text Available Neutrophils serve as a first line of defense in innate immunity owing in part to their ability to rapidly migrate towards chemotactic factors derived from invading pathogens. As a migratory function, neutrophil chemotaxis is regulated by the Rho family of small GTPases. However, the mechanisms by which Rho GTPases orchestrate cytoskeletal dynamics in migrating neutrophils remain ill-defined. In this study, we characterized the role of p21-activated kinase (PAK downstream of Rho GTPases in cytoskeletal remodeling and chemotactic processes of human neutrophils. We found that PAK activation occurred upon stimulation of neutrophils with f-Met-Leu-Phe (fMLP, and PAK accumulated at the actin-rich leading edge of stimulated neutrophils, suggesting a role for PAK in Rac-dependent actin remodeling. Treatment with the pharmacological PAK inhibitor, PF3758309, abrogated the integrity of RhoA-mediated actomyosin contractility and surface adhesion. Moreover, inhibition of PAK activity impaired neutrophil morphological polarization and directional migration under a gradient of fMLP, and was associated with dysregulated Ca(2+ signaling. These results suggest that PAK serves as an important effector of Rho-family GTPases in neutrophil cytoskeletal reorganization, and plays a key role in driving efficient directional migration of human neutrophils.

  10. Diagnostic system and diagnostic experiences at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Katona, Tamas

    1986-01-01

    The major functions of the diagnostic system of the first two units of the Paks Nuclear Power Plant are as follows: monitoring the mechanical integrity of the reactor and the primary coolant circuit by means of vibration diagnostics; leakage detection of the primary coolant circuit by means of high frequency sonic analysis; loose parts monitoring based on the analysis of high frequency signals of acceleration detectors; and monitoring the vibration state of the turbines and rotary machines by the latter method or by a procedure based on the detection of mechanical vibrations. Up-to-date vibration diagnostics is based on the information supplied by either acceleration detectors or pressure fluctuation detectors, or in-core and ex-core neutron detectors. (V.N.)

  11. GL-1196 Suppresses the Proliferation and Invasion of Gastric Cancer Cells via Targeting PAK4 and Inhibiting PAK4-Mediated Signaling Pathways

    Directory of Open Access Journals (Sweden)

    Jian Zhang

    2016-04-01

    Full Text Available Gastric cancer, which is the most common malignant gastrointestinal tumor, has jumped to the third leading cause of cancer-related mortality worldwide. It is of great importance to identify novel and potent drugs for gastric cancer treatment. P21-activated kinase 4 (PAK4 has emerged as an attractive target for the development of anticancer drugs in consideration of its vital functions in tumorigenesis and progression. In this paper, we reported that GL-1196, as a small molecular compound, effectively suppressed the proliferation of human gastric cancer cells through downregulation of PAK4/c-Src/EGFR/cyclinD1 pathway and CDK4/6 expression. Moreover, GL-1196 prominently inhibited the invasion of human gastric cancer cells in parallel with blockage of the PAK4/LIMK1/cofilin pathway. Interestingly, GL-1196 also inhibited the formation of filopodia and induced cell elongation in SGC7901 and BGC823 cells. Taken together, these results provided novel insights into the potential therapeutic strategy for gastric cancer.

  12. Isolation and Functional Characterization of a Floral Repressor, BcMAF1, From Pak-choi (Brassica rapa ssp. Chinensis).

    Science.gov (United States)

    Huang, Feiyi; Liu, Tongkun; Hou, Xilin

    2018-01-01

    MADS-box genes form a large gene family in plants and are involved in multiple biological processes, such as flowering. However, the regulation mechanism of MADS-box genes in flowering remains unresolved, especially under short-term cold conditions. In the present study, we isolated BcMAF1 , a Pak-choi ( Brassica rapa ssp. Chinensis ) MADS AFFECTING FLOWERING ( MAF ), as a floral repressor and functionally characterized BcMAF1 in Arabidopsis and Pak-choi. Subcellular localization and sequence analysis indicated that BcMAF1 was a nuclear protein and contained a conserved MADS-box domain. Expression analysis revealed that BcMAF1 had higher expression levels in leaves, stems, and petals, and could be induced by short-term cold conditions in Pak-choi. Overexpressing BcMAF1 in Arabidopsis showed that BcMAF1 had a negative function in regulating flowering, which was further confirmed by silencing endogenous BcMAF1 in Pak-choi. In addition, qPCR results showed that AtAP3 expression was reduced and AtMAF2 expression was induced in BcMAF1 -overexpressing Arabidopsis . Meanwhile, BcAP3 transcript was up-regulated and BcMAF2 transcript was down-regulated in BcMAF1 -silencing Pak-choi. Yeast one-hybrid and dual luciferase transient assays showed that BcMAF1 could bind to the promoters of BcAP3 and BcMAF2 . These results indicated that BcAP3 and BcMAF2 might be the targets of BcMAF1. Taken together, our results suggested that BcMAF1 could negatively regulate flowering by directly activating BcMAF2 and repressing BcAP3 .

  13. Review report on the dynamical study of the main building of the Paks NPP

    International Nuclear Information System (INIS)

    Gatti, F.

    1995-01-01

    The present report deals with the review of the report 'Dynamical Study of the main building of the Paks NPP', issued by Paks NPP (Hungary) on April, 1993, within the frame of the IAEA benchmark study for the seismic analysis and testing of an existing Nuclear Power Plant (M), and on behalf of ENEL DSR/VDN Rome, in the aims of the nuclear activities of ENEL DSR/VDN (Rome). After a foreword to define the aims of the job (Chapter 1) and the identification of the scope of the work (Chapter 2), a short list of references is given (Chapter 3). In Chapter 4, the criteria followed in the review activity are listed; in Chapter 5, the contents of the Paks NPP report are summarized. In Chapter 6 the results of the review are given, while the main conclusions of the review activities are summarized in the Chapter 7. (author)

  14. IAEA expert review mission completes assessment of fuel cleaning incident at Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    2003-01-01

    Full text: The IAEA today completed its expert review mission to investigate the 10 April fuel cleaning incident at the Paks nuclear power plant in Hungary. The mission was requested by the Hungarian Government to provide an independent assessment of the causes and actions taken by the plant and Hungarian authorities. The team was composed of nuclear and radiation experts from the IAEA, Austria, Canada, Finland, Slovakia, the United Kingdom and the United States. In a press conference, team leader Miroslav Lipar highlighted the team's findings in five areas: On management, the team concluded that the Hungarian Atomic Energy Authority and Paks are committed to improving the safety of the plant. They noted that as a result of steam generator decontamination in previous years, deposits became attached to the fuel assemblies. A decision was made to clean the fuel and contract an outside company to develop and operate a fuel cleaning process. The team found that the design and operation of the fuel cleaning tank and system was not accomplished in the manner prescribed by the IAEA Safety Standards. Neither the Hungarian Atomic Energy Authority nor Paks used conservative decision-making in their safety assessments for this unproven fuel cleaning system. The team determined that there was an over-reliance on the contractor that had been selected for the design, management and operation of the fuel cleaning system. Time pressure related to a prescribed fuel outage schedule, combined with confidence generated by previous successful fuel cleaning operations, contributed to a weak assessment of a new design and operation, which involved fuel directly removed from the reactor following a planned shutdown. On regulatory oversight, the IAEA team concluded that the Hungarian Atomic Energy Authority underestimated the safety significance of the proposed designs for the fuel cleaning system, which resulted in a less than rigorous review and assessment than should have been necessary

  15. Calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating water-cooled and moderated reactor WWER-type NPPs shake table investigation at Paks NPP. Final report from 15 June 1993 - 15 June 1994

    International Nuclear Information System (INIS)

    Kaznovsky, S.

    1995-01-01

    This final report involves the calculation and experimental examination and ensuring the seismic resistance of the reactor equipment and pipelines at start up and operation of WWER type nuclear power plants. Shake table experiments performed at the Paks NPP are included. Namely the following devices of the emergency cooling system were tested: pump of low pressure; valve of low pressure; intermediate heat exchanger. The following values were determined: natural frequencies and vibration decrements and the main modes of normal vibrations for the heat exchanger

  16. Glucosinolates from pak choi and broccoli induce enzymes and inhibit inflammation and colon cancer differently.

    Science.gov (United States)

    Lippmann, Doris; Lehmann, Carsten; Florian, Simone; Barknowitz, Gitte; Haack, Michael; Mewis, Inga; Wiesner, Melanie; Schreiner, Monika; Glatt, Hansruedi; Brigelius-Flohé, Regina; Kipp, Anna P

    2014-06-01

    High consumption of Brassica vegetables is considered to prevent especially colon carcinogenesis. The content and pattern of glucosinolates (GSLs) can highly vary among different Brassica vegetables and may, thus, affect the outcome of Brassica intervention studies. Therefore, we aimed to feed mice with diets containing plant materials of the Brassica vegetables broccoli and pak choi. Further enrichment of the diets by adding GSL extracts allowed us to analyze the impact of different amounts (GSL-poor versus GSL-rich) and different patterns (broccoli versus pak choi) of GSLs on inflammation and tumor development in a model of inflammation-triggered colon carcinogenesis (AOM/DSS model). Serum albumin adducts were analyzed to confirm the up-take and bioactivation of GSLs after feeding the Brassica diets for four weeks. In agreement with their high glucoraphanin content, broccoli diets induced the formation of sulforaphane-lysine adducts. Levels of 1-methoxyindolyl-3-methyl-histidine adducts derived from neoglucobrassicin were the highest in the GSL-rich pak choi group. In the colon, the GSL-rich broccoli and the GSL-rich pak choi diet up-regulated the expression of different sets of typical Nrf2 target genes like Nqo1, Gstm1, Srxn1, and GPx2. GSL-rich pak choi induced the AhR target gene Cyp1a1 but did not affect Ugt1a1 expression. Both colitis and tumor number were drastically reduced after feeding the GSL-rich pak choi diet while the other three diets had no effect. GSLs can act anti-inflammatory and anti-carcinogenic but both effects depend on the specific amount and pattern of GSLs within a vegetable. Thus, a high Brassica consumption cannot be generally considered to be cancer-preventive.

  17. MiR-145 regulates PAK4 via the MAPK pathway and exhibits an antitumor effect in human colon cells

    International Nuclear Information System (INIS)

    Wang, Zhigang; Zhang, Xiaoping; Yang, Zhili; Du, Hangxiang; Wu, Zhenqian; Gong, Jianfeng; Yan, Jun; Zheng, Qi

    2012-01-01

    Highlights: ► MiR-145 targets a putative binding site in the 3′UTR of PAK4. ► MiR-145 played an important role in inhibiting cell growth by directly targeting PAK4. ► MiR-145 may function as tumor suppressors. -- Abstract: MicroRNAs (miRNAs) are regulators of numerous cellular events; accumulating evidence indicates that miRNAs play a key role in a wide range of biological functions, such as cellular proliferation, differentiation, and apoptosis in cancer. Down-regulated expression of miR-145 has been reported in colon cancer tissues and cell lines. The molecular mechanisms underlying miR-145 and the regulation of colon carcinogenesis remain unclear. In this study, we investigated the levels of miR-145 in human colon cancer cells using qRT-PCR and found markedly decreased levels compared to normal epithelial cells. We identified PAK4 as a novel target of miR-145 using informatics screening. Additionally, we demonstrated that miR-145 targets a putative binding site in the 3′UTR of PAK4 and that its abundance is inversely associated with miR-145 expression in colon cancer cells; we confirmed this relationship using the luciferase reporter assay. Furthermore, restoration of miR-145 by mimics in SW620 cells significantly attenuated cell growth in vitro, in accordance with the inhibitory effects induced by siRNA mediated knockdown of PAK4. Taken together, these findings demonstrate that miR-145 downregulates P-ERK expression by targeting PAK4 and leads to inhibition of tumor growth.

  18. p21-activated Kinase1(PAK1) can promote ERK activation in a kinase independent manner

    DEFF Research Database (Denmark)

    Wang, Zhipeng; Fu, Meng; Wang, Lifeng

    2013-01-01

    204) although phosphorylation of b-Raf (Ser445) and c-Raf (Ser 338) remained unchanged. Furthermore, increased activation of the PAK1 activator Rac1 induced the formation of a triple complex of Rac1, PAK1 and Mek1, independent of the kinase activity of PAK1. These data suggest that PAK1 can stimulate...... MEK activity in a kinase independent manner, probably by serving as a scaffold to facilitate interaction of c-Raf....

  19. PLEX at Paks: making a virtue out of necessity

    International Nuclear Information System (INIS)

    Katona, T.; Bajsz, J.

    1992-01-01

    There are four VVER-440 units in operation at Paks Nuclear plant in Hungary. The units are of the V-213 type, ie the design which approaches the current, demanding Western standards in many aspects. During construction a number of innovations were adopted in these units. In particular, the instrumentation and control system was thoroughly improved, but there were also some changes to the main components (eg the pressure vessel). Although the quality assurance carried out during construction and commissioning was not the same as in Western practice, it was effective and resulted in relatively well constructed and tested units. Due to this, Pak has a good foundation on which to build a life extension programme-high availability, quality upgrading, a high integrity pressure vessel and a careful operating policy. Although life extension, economically speaking, is a necessity and not an option for Paks, the programme in itself should bring other benefits which would pay for themselves within the plant's design lifetime. (author)

  20. Lessons from the PAKS NPP case study

    International Nuclear Information System (INIS)

    Ronaky, J.

    2007-01-01

    A serious fuel cleaning incident happened in 2003 at the Hungarian Paks Nuclear Power Plant resulting in 30 damaged fuel bundles. The event was thoroughly investigated by the national authorities and reviewed by an IAFA team. Recovery operations have been successfully finished recently. The event attracted wide political and media coverage. Regulatory aspects of the event and the preparation for and realisation of the recovery operations will be presented with special emphasis on transparency and openness. Communication of the event itself and the national and international review process was challenging, but openness resulted in reconciliation of the Hungarian public. Recovery operations were accomplished after a careful preparation that took about three years. The situation was further complicated by the fact that the plant decided to start the operation or the reactor next to the cleaning tank before the recovery action. Some changes had to be licensed by the Regulatory Body in order to start the operation of the reactor. It attracted quite a big media interest. Detailed communication plans were prepared and followed both by the Regulatory Body and the Operator. Stakeholders were regularly invited to the plant to witness the operations and milestones of the process. NGOs requested the Regulatory Body to make public all technical data of both the operation of the reactor and the recovery process. Legal procedures in the court are going on to determine the extent and nature of data publicity associated with the recovery operations, while the Operator claims that technical details are proprietary information and not fully public. In the meantime lifetime extension of the plant and the construction of a low and intermediate level radioactive waste repository were debated and approved by the Hungarian Parliament. Good communication and open debate resulted in a wide political consensus and high public support in Hungary on the future of nuclear energy. (author)

  1. PAH analysis in Leipzig allotment soils; Untersuchungen zum Gefaehrdungspotential polycyclischer aromatischer Kohlenwasserstoffe (PAK) in Leipziger Kleingartenboeden

    Energy Technology Data Exchange (ETDEWEB)

    Bittrich, R.; Butze, B.; Mueller, S.; Prawalsky, R.; Stoye, H. [Umwelt-Consult e.V., Leipzig (Germany)

    2000-09-01

    Soils in 29 allotments were analyzed systematically with a view to the following aspects: Concentration ratios of the 16 components analyzed. Occurrence and classification of so-called PAH patterns. Interdependences between PAH patterns and soil features. PAH concentrations and soil-immanent buffer characteristics (humus concentration, pH, clay concentration, sesquioxide concentrations, exchange capacity). [German] Die vorliegende Arbeit konzentriert sich auf die Untersuchung der PAK-Belastung kleingaertnerisch genutzter Boeden. Die hier vorgestellten Ergebnisse resultieren aus Probjekten von Umwelt-Consult e.V. aus den Jahren 1995 bis 1997 im Auftrag der Stadt Leipzig und dem unter fachlicher Begleitung des Referates Geochemie der Abt. Boden/Geochemie vom LfUG gefoerderten Forschungsvorhaben 'Untersuchungen zum Gefaehrdungspotential polycyclischer aromatischer Kohlenwasserstoffe (PAK) in Boeden der Stadt Leipzig'. Hierbei wurden systematisch Boeden in 29 Kleingartenanlagen untersucht. Folgende Fragestellungen sollten beantwortet werden: Stehen die PAK-Konzentrationen der 16 analysierten Einzelkomponenten in bestimmten Groessenverhaeltnissen zueinander? Sind sogenannte PAK-Muster zu erkennen und lassen sich diese klassifizieren? Welche Beziehungen gibt es zwischen PAK-Mustern und Bodenmerkmalen? Korrespondieren die PAK-Konzentrationen (Gesamt-PAK, Einzelkomponenten) im Boden und deren bodenhorizont-bezogene Abfolge mit der Auspraegung bodenimmanenter Puffermerkmale (Humusgehalt, pH-Wert, Tongehalt, Gehalt an Sesquioxiden, Austauschkapazitaet)? (orig.)

  2. miR-155 Controls Lymphoproliferation in LAT Mutant Mice by Restraining T-Cell Apoptosis via SHIP-1/mTOR and PAK1/FOXO3/BIM Pathways.

    Directory of Open Access Journals (Sweden)

    Alexandre K Rouquette-Jazdanian

    Full Text Available Linker for Activation of T cells (LAT is an adapter protein that is essential for T cell function. Knock-in mice with a LAT mutation impairing calcium flux develop a fatal CD4+ lymphoproliferative disease. miR-155 is a microRNA that is correlated with hyperproliferation in a number of cancers including lymphomas and leukemias and is overexpressed in mutant LAT T cells. To test whether miR-155 was merely indicative of T cell activation or whether it contributes to lymphoproliferative disease in mutant LAT mice, we interbred LAT mutant and miR-155-deficient mice. miR-155 deficiency markedly inhibited lymphoproliferative disease by stimulating BIM-dependent CD4+ T cell apoptosis, even though ERK activation and T cell proliferation were increased in double mutant CD4+ T cells. Bim/Bcl2l11 expression is activated by the forkhead transcription factor FOXO3. Using miR-155-deficient, LAT mutant T cells as a discovery tool, we found two connected pathways that impact the nuclear translocation and activation of FOXO3 in T cells. One pathway is mediated by the inositide phosphatase SHIP-1 and the serine/threonine kinases AKT and PDK1. The other pathway involves PAK1 and JNK kinase activation. We define crosstalk between the two pathways via the kinase mTOR, which stabilizes PAK1. This study establishes a role for PAK1 in T cell apoptosis, which contrasts to its previously identified role in T cell proliferation. Furthermore, miR-155 regulates the delicate balance between PAK1-mediated proliferation and apoptosis in T cells impacting lymphoid organ size and function.

  3. An emerging role for p21-activated kinases (Paks) in viral infections

    DEFF Research Database (Denmark)

    Van den Broeke, Celine; Radu, Maria; Chernoff, Jonathan

    2010-01-01

    and motility, and abnormal Pak function is associated with a number of human diseases. Here, we discuss emerging evidence that these enzymes also play a major role in the entry, replication and spread of many important pathogenic human viruses, including HIV. Careful assessment of the potential role of Paks...

  4. MiR-145 regulates PAK4 via the MAPK pathway and exhibits an antitumor effect in human colon cells

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zhigang [Department of General Surgery, Shanghai Jiaotong University Affiliated 6th People' s Hospital, Shanghai (China); Zhang, Xiaoping [Department of Nuclear Medicine, Shanghai 10th People' s Hospital, Tongji University School of Medicine (China); Yang, Zhili; Du, Hangxiang; Wu, Zhenqian; Gong, Jianfeng; Yan, Jun [Department of General Surgery, Shanghai Jiaotong University Affiliated 6th People' s Hospital, Shanghai (China); Zheng, Qi, E-mail: zhengqi1957@yahoo.com.cn [Department of General Surgery, Shanghai Jiaotong University Affiliated 6th People' s Hospital, Shanghai (China)

    2012-10-26

    Highlights: Black-Right-Pointing-Pointer MiR-145 targets a putative binding site in the 3 Prime UTR of PAK4. Black-Right-Pointing-Pointer MiR-145 played an important role in inhibiting cell growth by directly targeting PAK4. Black-Right-Pointing-Pointer MiR-145 may function as tumor suppressors. -- Abstract: MicroRNAs (miRNAs) are regulators of numerous cellular events; accumulating evidence indicates that miRNAs play a key role in a wide range of biological functions, such as cellular proliferation, differentiation, and apoptosis in cancer. Down-regulated expression of miR-145 has been reported in colon cancer tissues and cell lines. The molecular mechanisms underlying miR-145 and the regulation of colon carcinogenesis remain unclear. In this study, we investigated the levels of miR-145 in human colon cancer cells using qRT-PCR and found markedly decreased levels compared to normal epithelial cells. We identified PAK4 as a novel target of miR-145 using informatics screening. Additionally, we demonstrated that miR-145 targets a putative binding site in the 3 Prime UTR of PAK4 and that its abundance is inversely associated with miR-145 expression in colon cancer cells; we confirmed this relationship using the luciferase reporter assay. Furthermore, restoration of miR-145 by mimics in SW620 cells significantly attenuated cell growth in vitro, in accordance with the inhibitory effects induced by siRNA mediated knockdown of PAK4. Taken together, these findings demonstrate that miR-145 downregulates P-ERK expression by targeting PAK4 and leads to inhibition of tumor growth.

  5. PAK4 crystal structures suggest unusual kinase conformational movements.

    Science.gov (United States)

    Zhang, Eric Y; Ha, Byung Hak; Boggon, Titus J

    2018-02-01

    In order for protein kinases to exchange nucleotide they must open and close their catalytic cleft. These motions are associated with rotations of the N-lobe, predominantly around the 'hinge region'. We conducted an analysis of 28 crystal structures of the serine-threonine kinase, p21-activated kinase 4 (PAK4), including three newly determined structures in complex with staurosporine, FRAX486, and fasudil (HA-1077). We find an unusual motion between the N-lobe and C-lobe of PAK4 that manifests as a partial unwinding of helix αC. Principal component analysis of the crystal structures rationalizes these movements into three major states, and analysis of the kinase hydrophobic spines indicates concerted movements that create an accessible back pocket cavity. The conformational changes that we observe for PAK4 differ from previous descriptions of kinase motions, and although we observe these differences in crystal structures there is the possibility that the movements observed may suggest a diversity of kinase conformational changes associated with regulation. Protein kinases are key signaling proteins, and are important drug targets, therefore understanding their regulation is important for both basic research and clinical points of view. In this study, we observe unusual conformational 'hinging' for protein kinases. Hinging, the opening and closing of the kinase sub-domains to allow nucleotide binding and release, is critical for proper kinase regulation and for targeted drug discovery. We determine new crystal structures of PAK4, an important Rho-effector kinase, and conduct analyses of these and previously determined structures. We find that PAK4 crystal structures can be classified into specific conformational groups, and that these groups are associated with previously unobserved hinging motions and an unusual conformation for the kinase hydrophobic core. Our findings therefore indicate that there may be a diversity of kinase hinging motions, and that these may

  6. The Hungarian model project: Strengthening training for operational safety at Paks nuclear power plant

    International Nuclear Information System (INIS)

    Mautner Markhof, F.

    1998-01-01

    The Hungarian Model project (HMP) reflects the commitment to constant increase of safety and reliability of the NPP Paks, the Government of Hungary and the IAEA. It includes some of the most important nuclear power objectives of Paks NPP, namely the strengthening of NPP personnel training and competence through the application of international best practice, the systematic approach to training (SAT), for training operation and maintenance personnel; setting up a state of-the-art maintenance training center (MTC) at Paks and enhancing safety culture at Paks NPP. The IAEA supported implementation of the HMP through fellowships and scientific visits, expert missions, provision of hardware and software for SAT application, and supply od major new uncontaminated items of actual WWER equipment for the MTC

  7. Full-scale dynamic structural testing of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Da Rin, E.M.; Muzzi, F.P.

    1995-01-01

    Within the framework of the IAEA coordinated 'Benchmark Study for the seismic analysis and testing of WWER-type NPPs', in-situ dynamic structural testing activities have been performed at the Paks Nuclear Power Plant in Hungary. The specific objective of the investigation was to obtain experimental data on the actual dynamic structural behaviour of the plant's major constructions and equipment under normal operating conditions, for enabling a valid seismic safety review to be made. This paper gives a synthetic description of the conducted experiments and presents some results, regarding in particular the free-field excitations produced during the earthquake-simulation experiments and an experiment of the dynamic soil-structure interaction global effects at the base of the reactor containment structure. Moreover, a method which can be used for inferring dynamic structural characteristics from the recorded time-histories is briefly described and a simple illustrative example given. (author)

  8. Cloning and characterization of PAK5, a novel member of mammalian p21-activated kinase-II subfamily that is predominantly expressed in brain

    DEFF Research Database (Denmark)

    Pandey, A.; Dan, I.; Kristiansen, T.Z.

    2002-01-01

    cloned a novel human PAK family kinase that has been designated as PAK5. PAK5 contains a CDC42/Rac1 interactive binding (CRIB) motif at the N-terminus and a Ste20-like kinase domain at the C-terminus. PAK5 is structurally most related to PAK4 and PAK6 to make up the PAK-II subfamily. We have shown...

  9. Summary of IVO participation in Paks blast test analysis

    International Nuclear Information System (INIS)

    Varpasuo, P.

    2001-01-01

    The paper deals with the numerical simulation of the triple blast test performed at Paks NPP. A detailed background analysis was carried out to complete the geological and geotechnical properties and, consequently, special frequency dependent soil stiffnesses have been evaluated. The structural model (3D) allowed a very refined result presentation in terms of profiles of displacements and forces at different elevations, for direct comparison with the experimental output. (author)

  10. Nonautonomous Regulation of Neuronal Migration by Insulin Signaling, DAF-16/FOXO, and PAK-1

    Directory of Open Access Journals (Sweden)

    Lisa M. Kennedy

    2013-09-01

    Full Text Available Neuronal migration is essential for nervous system development in all organisms and is regulated in the nematode, C. elegans, by signaling pathways that are conserved in humans. Here, we demonstrate that the insulin/IGF-1-PI3K signaling pathway modulates the activity of the DAF-16/FOXO transcription factor to regulate the anterior migrations of the hermaphrodite-specific neurons (HSNs during embryogenesis of C. elegans. When signaling is reduced, DAF-16 is activated and promotes migration; conversely, when signaling is enhanced, DAF-16 is inactivated, and migration is inhibited. We show that DAF-16 acts nonautonomously in the hypodermis to promote HSN migration. Furthermore, we identify PAK-1, a p21-activated kinase, as a downstream mediator of insulin/IGF-1-DAF-16 signaling in the nonautonomous control of HSN migration. Because a FOXO-Pak1 pathway was recently shown to regulate mammalian neuronal polarity, our findings indicate that the roles of FOXO and Pak1 in neuronal migration are most likely conserved from C. elegans to higher organisms.

  11. An AFLP marker linked to turnip mosaic virus resistance gene in pak ...

    African Journals Online (AJOL)

    An AFLP marker linked to turnip mosaic virus resistance gene in pak-choi. W Xinhua, C Huoying, Z Yuying, H Ruixian. Abstract. Pak-choi is one of the most important vegetable crops in China. Turnip mosaic virus (TuMV) is one of its main pathogen. Screening the molecular marker linked to the TuMV resistance gene is an ...

  12. Surveillance extension experience at WWER-440 type reactors

    International Nuclear Information System (INIS)

    Gillemot, F.; Uri, G.; Oszwald, F.; Trampus, P.

    1993-01-01

    In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs

  13. Surveillance extension experience at WWER-440 type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gillemot, F; Uri, G [Budapesti Mueszaki Egyetem, Budapest (Hungary); Oszwald, F; Trampus, P

    1994-12-31

    In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs.

  14. Hair Growth Promoting and Anticancer Effects of p21-activated kinase 1 (PAK1 Inhibitors Isolated from Different Parts of Alpinia zerumbet

    Directory of Open Access Journals (Sweden)

    Nozomi Taira

    2017-01-01

    Full Text Available PAK1 (p21-activated kinase 1 is an emerging target for the treatment of hair loss (alopecia and cancer; therefore, the search for PAK1 blockers to treat these PAK1-dependent disorders has received much attention. In this study, we evaluated the anti-alopecia and anticancer effects of PAK1 inhibitors isolated from Alpinia zerumbet (alpinia in cell culture. The bioactive compounds isolated from alpinia were found to markedly promote hair cell growth. Kaempferol-3-O-β-d-glucuronide (KOG and labdadiene, two of the isolated compounds, increased the proliferation of human follicle dermal papilla cells by approximately 117%–180% and 132%–226%, respectively, at 10–100 μM. MTD (2,5-bis(1E,3E,5E-6-methoxyhexa-1,3,5-trien-1-yl-2,5-dihydrofuran and TMOQ ((E-2,2,3,3-tetramethyl-8-methylene-7-(oct-6-en-1-yloctahydro-1H-quinolizine showed growth-promoting activity around 164% and 139% at 10 μM, respectively. The hair cell proliferation induced by these compounds was significantly higher than that of minoxidil, a commercially available treatment for hair loss. Furthermore, the isolated compounds from alpinia exhibited anticancer activity against A549 lung cancer cells with IC50 in the range of 67–99 μM. Regarding the mechanism underlying their action, we hypothesized that the anti-alopecia and anticancer activities of these compounds could be attributed to the inhibition of the oncogenic/aging kinase PAK1.

  15. Waste Cellulose from Tetra Pak Packages as Reinforcement of Cement Concrete

    Directory of Open Access Journals (Sweden)

    Gonzalo Martínez-Barrera

    2015-01-01

    Full Text Available The development of the packaging industry has promoted indiscriminately the use of disposable packing as Tetra Pak, which after a very short useful life turns into garbage, helping to spoil the environment. One of the known processes that can be used for achievement of the compatibility between waste materials and the environment is the gamma radiation, which had proved to be a good tool for modification of physicochemical properties of materials. The aim of this work is to study the effects of waste cellulose from Tetra Pak packing and gamma radiation on the mechanical properties of cement concrete. Concrete specimens were elaborated with waste cellulose at concentrations of 3, 5, and 7 wt% and irradiated at 200, 250, and 300 kGy of gamma dose. The results show highest improvement on the mechanical properties for concrete with 3 wt% of waste cellulose and irradiated at 300 kGy; such improvements were related with the surface morphology of fracture zones of cement concrete observed by SEM microscopy.

  16. External hazards considered for Paks NPP

    International Nuclear Information System (INIS)

    Kiss, Tibor

    2000-01-01

    PAKS NPP was built according to Soviet construction standards which took into account meteorological aspects but no documents for other external hazards were available. Main activities concerning earthquakes cover reevaluation of the plant site, seismic safety technological concept, improving the seismic resistance, installation of seismic monitoring and protection system, and seismic PSA

  17. Pertumbuhan Dan Produktivitas Sawi Pak Choy (Brasica Rapa L.) Pada Umur Transplanting Dan Pemberian Mulsa Organik

    OpenAIRE

    Pribadi, Gandhi Yudhistira; Roviq, Mochammad; Wardiyati, Tatik

    2014-01-01

    Potensi produksi tanaman pak choy belum optimal, rendahnya produksi pak choy dikarenakan pada teknik budidayanya petani cendrung tidak memperhatikan kondisi lingkungan mikro dan masih belum adanya standart transplanting yang tepat. Penelitian bertujuan untuk mendapatkan teknik budidaya pak choy dengan penggunaan mulsa dan saat transplanting yang tepat. Dilaksanakan pada bulan Mei - Juli 2013 di Desa Pandanrejo, Kecamatan Bumiaji - Batu. Penelitian menggunakan Ranca-ngan Acak Kelompok Faktoria...

  18. Reciprocally coupled residues crucial for protein kinase Pak2 activity calculated by statistical coupling analysis.

    Directory of Open Access Journals (Sweden)

    Yuan-Hao Hsu

    2010-03-01

    Full Text Available Regulation of Pak2 activity involves at least two mechanisms: (i phosphorylation of the conserved Thr(402 in the activation loop and (ii interaction of the autoinhibitory domain (AID with the catalytic domain. We collected 482 human protein kinase sequences from the kinome database and globally mapped the evolutionary interactions of the residues in the catalytic domain with Thr(402 by sequence-based statistical coupling analysis (SCA. Perturbation of Thr(402 (34.6% suggests a communication pathway between Thr(402 in the activation loop, and Phe(387 (DeltaDeltaE(387F,402T = 2.80 in the magnesium positioning loop, Trp(427 (DeltaDeltaE(427W,402T = 3.12 in the F-helix, and Val(404 (DeltaDeltaE(404V,402T = 4.43 and Gly(405 (DeltaDeltaE(405G,402T = 2.95 in the peptide positioning loop. When compared to the cAMP-dependent protein kinase (PKA and Src, the perturbation pattern of threonine phosphorylation in the activation loop of Pak2 is similar to that of PKA, and different from the tyrosine phosphorylation pattern of Src. Reciprocal coupling analysis by SCA showed the residues perturbed by Thr(402 and the reciprocal coupling pairs formed a network centered at Trp(427 in the F-helix. Nine pairs of reciprocal coupling residues crucial for enzymatic activity and structural stabilization were identified. Pak2, PKA and Src share four pairs. Reciprocal coupling residues exposed to the solvent line up as an activation groove. This is the inhibitor (PKI binding region in PKA and the activation groove for Pak2. This indicates these evolutionary conserved residues are crucial for the catalytic activity of PKA and Pak2.

  19. Experience in modernization of safety I and C in VVER 440 nuclear power plants Bohunice V1 and Paks

    International Nuclear Information System (INIS)

    Martin, M.

    2000-01-01

    For nuclear power plants which have been in operation for more than 15 years, backfitting or even complete replacement of the instrumentation and control (I and C) equipment becomes an increasingly attractive option, motivated not only by the objective to reduce the cost of I and C system maintenance and repair but also to prolong or at least to safeguard the plant life-time: optimized spare-part management through use of standard equipment; reduction of number and variety of different items of equipment by implementing functions stepwise in application software; adding new functionality in the application software which was not possible in the old technology due to lack of space; safeguarding of long-term After-Sales-Service. Some years ago Bohunice V1 NPP, Slovak Republic and Paks NPP, Hungary intended to replace parts of their Safety I and C, mainly the Reactor Trip System, the Reactor Limitation System and the Neutron Flux Excore Instrumentation and Monitoring Systems. After a Basic Engineering Phase in Bohunice V1 and a Feasibility Study in Paks both plants decided to use the Siemens' Digital Safety I and C System TELEPERM XS to modernize their plants. Both Bohunice, Unit 2 and Paks, Unit 1 have been back on line for over six months with the new Digital Safety I and C. At the present time Bohunice, Unit 1 and within the next few months Paks, Unit 2 will be replaced. Trouble-free start-ups and no major problems under operation in the first two plants were based on: thorough understanding of the VVER 440 technology; comprehensive planning together with the plant operators and authorities; the possibility to adapt TELEPERM XS to every plant type; the execution of extensive pre-operational tests. Regarding these modernization measures Siemens, as well as the above Operators, have gained considerable experience in the field of I and C Modernization in VVER 440 NPPs. Important aspects of this experience are: how to transfer the VVER technology to TELEPERM XS; how to

  20. Radiation protection aspects of the repair work at Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Bujtas, T.; Nenyei, A.

    2006-01-01

    On the Unit 2 at Paks Nuclear Power Plant accident occurred on 10th April 2003. Thirty fuel assemblies damaged in the cleaning tank installed in the Pit No. 1. Due to the accident casing of the fuel elements and uranium-dioxide pellets inside them damaged. The scratched fuel assemblies and nuclear fuel fragments should be removed and safely deposited. In order to restore the operational condition of the Pit No. 1 a lot of complicated activities with radiation hazard should be implemented. These tasks bring up both technical difficulties and serious radiation protection problems, and it is essential to resolve them in order to reduce radiation exposure of the working personnel and to minimize the amount of off-site radioactive releases.There was a serious incident (An INES level 3 event) at Paks Nuclear Power plant in april 10, 2003. (TRA)

  1. Non-autonomous Regulation of Neuronal Migration by Insulin Signaling, DAF-16/FOXO and PAK-1

    Science.gov (United States)

    Kennedy, Lisa M.; Pham, Steven C.D.L.; Grishok, Alla

    2013-01-01

    SUMMARY Neuronal migration is essential for nervous system development in all organisms and is regulated in the nematode, C. elegans, by signaling pathways that are conserved in humans. Here, we demonstrate that the Insulin/IGF-1-PI3K signaling pathway modulates the activity of the DAF-16/FOXO transcription factor to promote the anterior migrations of the hermaphrodite-specific neurons (HSNs) during embryogenesis of C. elegans. When signaling is reduced, DAF-16 is activated and promotes migration, conversely, when signaling is enhanced, DAF-16 is inactivated and migration is inhibited. We show that DAF-16 acts non-autonomously in the hypodermis to promote HSN migration. Furthermore, we identify PAK-1, a p21-activated kinase, as a downstream mediator of Insulin/IGF-1-DAF-16 signaling in the non-autonomous control of HSN migration. As a FOXO-Pak1 pathway was recently shown to regulate mammalian neuronal polarity, our findings indicate that the roles of FOXO and Pak1 in neuronal migration are likely conserved from C. elegans to higher organisms. PMID:23994474

  2. Stimulus-dependent regulation of the phagocyte NADPH oxidase by a VAV1, Rac1, and PAK1 signaling axis

    DEFF Research Database (Denmark)

    Roepstorff, Kirstine; Rasmussen, Izabela Zorawska; Sawada, Makoto

    2008-01-01

    dominant-positive mutants enhanced, whereas dominant-negative mutants inhibited, NADPH oxidase-mediated superoxide generation following formyl-methionyl-leucylphenylalanine or phorbol 12-myristate 13-acetate stimulation. Both Rac1 and the GTP exchange factor VAV1 were required as upstream signaling......The p21-activated kinase-1 (PAK1) is best known for its role in the regulation of cytoskeletal and transcriptional signaling pathways. We show here in the microglia cell line Ra2 that PAK1 regulates NADPH oxidase (NOX-2) activity in a stimulus-specific manner. Thus, conditional expression of PAK1...... proteins in the formyl-methionyl-leucyl-phenylalanine-induced activation of endogenous PAK1. In contrast, PAK1 mutants had no effect on superoxide generation downstream of FcgammaR signaling during phagocytosis of IgG-immune complexes. We further present evidence that the effect of PAK1 on the respiratory...

  3. Mobilization of PAH by synthetic gastrointestinal juice from contaminated soil of a former landfill area; Mobilisierung von PAK durch synthetische Verdauungssaefte aus dem kontaminierten Bodenmaterial einer Altlastenflaeche

    Energy Technology Data Exchange (ETDEWEB)

    Hack, A.; Selenka, F.; Wilhelm, M. [Bochum Univ. (Germany). Abt. fuer Hygiene, Sozial- und Umweltmedizin

    1998-10-01

    In the present study, the amount of polycyclic aromatic hydrocarbons (PAH) in contaminated soil material, which may be available for absorption in the gastrointestinal tract, is estimated by means of evaluating the PAH mobilization by synthetic gastric and intestinal juice in an in vitro test system. Five contaminated soil materials from a former landfill site are analysed in this gastrointestinal model for the PAH of the U.S.EPA-standard. For quantification, an HPLC method with reversed-phase chromatography and on line fluorescence detection is used. The PAH concentration of the contaminated soil materials ranged from 37 {mu}g/g up to 196 {mu}g/g in total. The mobilization of the PAH in the gastrointestinal model ranged from 0.3% up to 1.3% when gastrointestinal juice was used alone. In the presence of whole milk powder, however, the mobilization was enhanced to values from 10.8% up to 14.5%. Since the soil material was taken from different parts of the contaminated area, and since the mobilization of the PAH from the different materials shows only minor differences, the mobilization data evaluated may be considered as representative for the whole contaminated area. Compared to other contaminated soil materials, especially those from gas work areas or coke plants, the mobilization rate of PAH by the gastrointestinal model from the soil materials used in this study is low. The health risk caused, by ingestion of this soil material, as far as PAH are concerned, is actually smaller than the risk calculated from the total content of PAH of the contaminated soil. (orig.) [Deutsch] Im allgemeinen wird nur ein Teil der Schadstoffe aus oral aufgenommenem kontaminiertem Bodenmaterial im Gastrointestinaltrakt resorbiert. In der vorliegenden Studie wird der resorptionsverfuegbare Anteil der PAK aus dem real kontaminierten Bodenmaterial einer ehemaligen Deponie aus dem sueddeutschen Raum anhand der Mobilisierbarkeit der PAK durch die Verdauungssaefte des oberen

  4. Experiences with the upgraded SKP system during refuelling Paks nuclear power plant

    International Nuclear Information System (INIS)

    Baranyai, A.; Hetzmann, A.

    1997-01-01

    In order to control the neutron flux during the refueling period, new measuring chains were developed and put into operation by the experts of KFKI-RegTron Co., Ltd. and the Paks Nuclear Power Plant with the purpose of partially substituting the original Refuelling Neutron Monitoring system (SKP) of WWER-440 reactor units. The modified monitoring system processes the signals of detectors located in channels outside the core. The outputs of measurement amplifiers equipped with up-to-date electronics fit in the original system perfectly. Use of the out-of-core measuring technique confirmed the preliminary expectations: interference sensitivity has decreased, the neutron/gamma ration increased and refueling time has become shorter by one to one-and-a-half day. The paper details the reasons for upgrading, the essence of utilized solutions and the operational experience. (author)

  5. Experience of Hungarian model project: 'Strengthening training for operational safety at Paks NPP'

    International Nuclear Information System (INIS)

    Kiss, I.

    1998-01-01

    Training of Operational Safety at Paks NPP is described including all the features of the project including namely: description of Paks NPP, its properties and performances; reasons for establishing Hungarian Model Project, its main goals, mentioning Hungarian and IAEA experts involved in the Project, its organization, operation, budget, current status together with its short term and long term impact

  6. DESIGNAND EVALUATION OF TETRA-PAK CONTAINERS' RECYCLING PLANT ON A SMALLSCALE

    OpenAIRE

    Inche Mitma, Jorge; Vergiú Canto, Jorge|; Mavila Hinojoza, Daniel; Godoy Martínez, Manuel; Chung Pinzás, Alfonso

    2014-01-01

    This study deals about the design and evaluation of Tetra Pak containers' recycling plant on a small scale. The basic Plant Engineering was found from the information gathered. Some aspects included were: product design, process design, equipment design and costs evaluation, with the aim of determining its technical, economical and environmental capability for its implementation. El estudio trata sobre el diseño y evaluación de una planta de reciclaje de envases tetra pak a pequeña escala....

  7. Strategi Pengembangan Usaha Ayam Potong Pak Imanto Di Pasar Tapiv Binjai

    OpenAIRE

    Meliala, Karina Octavina

    2014-01-01

    Chicken meat were popular among people. There are many reasons why people’s interest on buying chicken meat were high.Until this day, many people fond of chickend meat and that is what making this chicken meat business promising.Huge market makes this business promising in the future.One small scale enterpreneur who work in this business is Pak Imanto who opened Pak Imanto’s Chicken Meat retailer. In order to grow, this business needs to develop its streght, weakneses, opportunities and threa...

  8. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4F. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  9. Safety upgrading at PAKS Nuclear Power Plant

    International Nuclear Information System (INIS)

    Bajsz, J.; Elter, J.

    2000-01-01

    The operation of Paks NPP has reached its half time. Until this time the plant fulfilled expectations raised before its construction: the four units have produced safely and reliably more than 200 TWh electricity. The production of the plant has been at the stable level since its construction and has provided 43-38 % of electricity consumed in Hungary. The annual production is around 14 TWh, which means a load factor higher than 85 %. Safety upgrading activities [1] at Paks had started in the late eighties, when the commissioning work of units 3 and 4 were carried out. That time the main emphases were put to lessons learned of the TMI and Chernobyl accidents. The international reviews hosted by our plant widened our review's scope. To systematize our approach a complete safety review, the AGNES (Advanced General Safety and New Evaluation of Safety) project was started in 1991. The goal of the project was to evaluate to what extent Paks NPP satisfied the current international safety expectations and to help in determining the priorities for safety enhancement and upgrading measures. The project completed in 1994 ranked our safety upgrading measures by safety significance, which became a basis for technical design work and financial scheduling. The other important outcome of the AGNES project was the introduction the Periodical Safety Review regime by our nuclear authority. These periodical reviews held after 10 years of operation offer the possibility - and obligation for the licensee - to perform a comprehensive assessment of the safety of the plant, to evaluate the integral effects of changes of circumstances happened during the review period. The goal of these reviews is to deal with cumulative effects of NPP ageing, modifications, operating experience and technical developments aimed at ensuring a high level of safety throughout plant service life. The execution of our safety-upgrading program is well advancing. For the whole program from 1996 to 2002 250

  10. Database for the OECD-IAEA Paks Fuel Project

    International Nuclear Information System (INIS)

    Szabo, Emese; Hozer, Zoltan; Gyori, Csaba; Hegyi, Gyoergy

    2010-01-01

    On 10 April 2003 severe damage of fuel assemblies took place during an incident at Unit 2 of Paks Nuclear Power Plant in Hungary. The assemblies were being cleaned in a special tank below the water level of the spent fuel storage pool in order to remove crud buildup. That afternoon, the chemical cleaning of assemblies was completed and the fuel rods were being cooled by circulation of storage pool water. The first sign of fuel failure was the detection of some fission gases released from the cleaning tank during that evening. The cleaning tank cover locks were released after midnight and this operation was followed by a sudden increase in activity concentrations. The visual inspection revealed that all 30 fuel assemblies were severely damaged. The first evaluation of the event showed that the severe fuel damage happened due to inadequate coolant circulation within the cleaning tank. The damaged fuel assemblies will be removed from the cleaning tank in 2005 and will be stored in special canisters in the spent fuel storage pool of the Paks NPP. Following several discussions between expert from different countries and international organisations the OECD-IAEA Paks Fuel Project was proposed. The project is envisaged in two phases. - Phase 1 is to cover organization of visual inspection of material, preparation of database, performance of analyses and preparatory work for fuel examination. - Phase 2 is to cover the fuel transport and the hot cell examination. The first meeting of the project was held in Budapest on 30-31 January 2006. Phase 1 of the Paks Fuel Project will focus on the numerical simulation of the most important aspects of the incident. This activity will help in the reconstruction of the accidental scenario. The first step of Phase 1 was the collection of a database necessary for the code calculations. The main objective of database collection was to provide input data for calculations. For this reason the collection was focused on such data that are

  11. Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Baji, C.; Vamos, G.; Toth, J.

    2001-01-01

    The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

  12. The RELAP5-Based NPA of the VVER Type Paks NPP

    International Nuclear Information System (INIS)

    Guba, A.; Toth, I.; Mandy, C.; Stubbe, E.

    1999-01-01

    NPA is a data driven interactive graphical tool for visualisation of different plant conditions. Data generated by the analysis code RELAP5/MOD3.2 are processed and displayed on a computer monitor. The NPA model of Paks NPP Unit 3 was developed with the aim to demonstrate the phenomena occurring in different transient/accident scenarios. This VVER-specific NPA development is a result of a cooperation between BELGATOM and KFKI-AEKI. (author)

  13. Potential role of p21 Activated Kinase 1 (PAK1) in the invasion and motility of oral cancer cells

    International Nuclear Information System (INIS)

    Parvathy, Muraleedharan; Sreeja, Sreeharshan; Kumar, Rakesh; Pillai, Madhavan Radhakrishna

    2016-01-01

    Oral cancer malignancy consists of uncontrolled division of cells primarily in and around the floor of the oral cavity, gingiva, oropharynx, lower lip and base of the tongue. According to GLOBOCAN 2012 report, oral cancer is one of the most common cancers among males and females in India. Even though significant advancements have been made in the field of oral cancer treatment modalities, the overall prognosis for the patients has not improved in the past few decades and hence, this demands a new thrust for the identification of novel therapeutic targets in oral cancer. p21 Activated Kinases (PAKs) are potential therapeutic targets that are involved in numerous physiological functions. PAKs are serine-threonine kinases and they serve as important regulators of cytoskeletal dynamics and cell motility, transcription through MAP kinase cascades, death and survival signalling, and cell-cycle progression. Although PAKs are known to play crucial roles in cancer progression, the role and clinical significance of PAKs in oral cancer remains poorly understood. Our results suggest that PAK1 is over-expressed in oral cancer cell lines. Stimulation of Oral Squamous Cell Carcinoma (OSCC) cells with serum growth factors leads to PAK1 re-localization and might cause a profound cytoskeletal remodelling. PAK1 was also found to be involved in the invasion, migration and cytoskeletal remodelling of OSCC cells. Our study revealed that PAK1 may play a crucial role in the progression of OSCC. Studying the role of PAK1 and its substrates is likely to enhance our understanding of oral carcinogenesis and potential therapeutic value of PAKs in oral cancer. The online version of this article (doi:10.1186/s12885-016-2263-8) contains supplementary material, which is available to authorized users

  14. The maintenance training center of the paks nuclear power plant - past, present and future

    International Nuclear Information System (INIS)

    Kiss, I.

    2001-01-01

    The safety of the Paks nuclear power plant (Paks NPP) is a political-economic factor with general influence on the stability of the Hungarian economy. Since the beginning of the 1990s, the plant management has been taking significant efforts to learn about the factors that define plant safety and to reveal areas where safety can be further improved. Major emphasis is also placed on the provision of resources and creation of conditions necessary for the preservation of staff competence. In 1997 a separate, maintenance-specific facility was erected. The Maintenance Training Center of the Paks NPP is a worldwide major unique center. (orig.) [de

  15. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4B. Paks NPP: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on dynamic study of the main building of the Paks NPP; shake table investigation at Paks NPP and the Final report of the Co-ordinated Research Programme

  16. Interim storage of spent fuel elements in the Paks Nuclear Power Plant, Hungary

    International Nuclear Information System (INIS)

    Szabo, B.

    1998-01-01

    The interim storage of spent fuel cassettes of the Paks NPP provides storage for 50 years at the Paks NPP site. The modular dry storage technology is presented. The technological design and the licensing of the facility has been made by the GEC Alsthom ESL firm. This storage facility can accommodate 450 fuel cassettes until their final disposal. (R.P.)

  17. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4A. Paks NPP: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to seismic analyses of structures of Paks and Kozloduy reactor buildings and WWER-440/213 primary coolant loops with different antiseismic devices

  18. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4A. Paks NPP: Analysis/testing. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to seismic analyses of structures of Paks and Kozloduy reactor buildings and WWER-440/213 primary coolant loops with different antiseismic devices.

  19. Maintenance training centre at NPP Paks, Hungary

    International Nuclear Information System (INIS)

    Babos, K.

    1996-01-01

    The lecture shows the feature of WWER-440/213 units maintenance, the existing maintenance training system, the necessity of the change in maintenance training system at NPP Paks. The author introduces the would-be maintenance training centre, the training facilities and the main tasks related to the maintenance training. (author)

  20. Achievements and challenges of Paks NPP

    International Nuclear Information System (INIS)

    Bajsz, J.; Katona, T.

    2002-01-01

    As the six year long safety upgrading program at Paks NPP is approaching its final stage this year, it is a good opportunity to draw the conclusion: what have been done and how have measures influenced the safety of the plant. In its first part the paper gives an overview of the program's main issues, assesses the results from the point of view of safety, reliability and cost effectiveness as well. In the second part a survey of future tasks follows: (1) Hungary is joining to the EU. The accession process so far has not revealed any major problems concerning nuclear safety which could be seen as obstacles toward the membership. However the plant should be ready to meet the increasing level of safety regulations. Further safety upgrading measures are planned, mostly in the field of severe accident management. (2) The electricity market liberalisation in Hungary will start in 2003 and being a EU member state, the full market opening will happen within a few years. The plant has to take into account the specificity of market functioning. The most important thing is to preserve the present cost advantage of nuclear electricity generation within the market conditions. The paper presents measures performed and planned to keep the unit generation cost competitive. (3) The first unit at Paks will mark its 20't'h anniversary this year. Lifetime management issues are at the centre of the engineering activities. The work already started to prepare the lifetime extension for 20 years. The program for the license renewal, which was elaborated jointly with the nuclear regulatory body will be described.(author)

  1. A βPIX-PAK2 complex confers protection against Scrib-dependent and cadherin-mediated apoptosis

    DEFF Research Database (Denmark)

    Frank, Scott R; Bell, Jennifer H; Frödin, Morten

    2012-01-01

    During epithelial morphogenesis, a complex comprising the βPIX (PAK-interacting exchange factor β) and class I PAKs (p21-activated kinases) is recruited to adherens junctions. Scrib, the mammalian ortholog of the Drosophila polarity determinant and tumor suppressor Scribble, binds βPIX directly. ...

  2. Training system enhancement for nuclear safety at PAKS NPP

    International Nuclear Information System (INIS)

    KIss, I.

    2000-01-01

    Paks Nuclear Power Plant is the only commercial nuclear facility in Hungary, which has been operational since 1982. The over 15 years operation of the plant can from all aspects be considered as a success, to which the well qualified, competent staff significantly contributes. Like other N-plants, Paks NPP is also exposed to major challenges due to plant ageing and changes in circumstances that affect the operation. The management focusing on maintaining nuclear safety launched an overall programme to upgrade quality of personnel training and to improve its infrastructure. Though this programme has successfully finished with visible proofs, further actions to develop a reconsidered human resource policy is needed so that the plant would successfully stand against the challenges of the 21. century. (author)

  3. Erk5 inhibits endothelial migration via KLF2-dependent down-regulation of PAK1.

    Science.gov (United States)

    Komaravolu, Ravi K; Adam, Christian; Moonen, Jan-Renier A J; Harmsen, Martin C; Goebeler, Matthias; Schmidt, Marc

    2015-01-01

    The MEK5/Erk5 pathway mediates beneficial effects of laminar flow, a major physiological factor preventing vascular dysfunction. Forced Erk5 activation induces a protective phenotype in endothelial cell (EC) that is associated with a dramatically decreased migration capacity of those cells. Transcriptional profiling identified the Krüppel-like transcription factors KLF2 and KLF4 as central mediators of Erk5-dependent gene expression. However, their downstream role regarding migration is unclear and relevant secondary effectors remain elusive. Here, we further investigated the mechanism underlying Erk5-dependent migration arrest in ECs. Our experiments reveal KLF2-dependent loss of the pro-migratory Rac/Cdc42 mediator, p21-activated kinase 1 (PAK1), as an important mechanism of Erk5-induced migration inhibition. We show that endothelial Erk5 activation by expression of a constitutively active MEK5 mutant, by statin treatment, or by application of laminar shear stress strongly decreased PAK1 mRNA and protein expression. Knockdown of KLF2 but not of KLF4 prevented Erk5-mediated PAK1 mRNA inhibition, revealing KLF2 as a novel PAK1 repressor in ECs. Importantly, both PAK1 re-expression and KLF2 knockdown restored the migration capacity of Erk5-activated ECs underscoring their functional relevance downstream of Erk5. Our data provide first evidence for existence of a previously unknown Erk5/KLF2/PAK1 axis, which may limit undesired cell migration in unperturbed endothelium and lower its sensitivity for migratory cues that promote vascular diseases including atherosclerosis. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2014. For permissions please email: journals.permissions@oup.com.

  4. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4E. Paks NPP: Analysis and testing. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  5. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4E. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  6. Lentiviral Nef Proteins Utilize PAK2-Mediated Deregulation of Cofilin as a General Strategy To Interfere with Actin Remodeling▿ †

    Science.gov (United States)

    Stolp, Bettina; Abraham, Libin; Rudolph, Jochen M.; Fackler, Oliver T.

    2010-01-01

    Nef is an accessory protein and pathogenicity factor of human immunodeficiency virus (HIV) and simian immunodeficiency virus (SIV) which elevates virus replication in vivo. We recently described for HIV type 1SF2 (HIV-1SF2) the potent interference of Nef with T-lymphocyte chemotaxis via its association with the cellular kinase PAK2. Mechanistic analysis revealed that this interaction results in deregulation of the actin-severing factor cofilin and thus blocks the chemokine-mediated actin remodeling required for cell motility. However, the efficiency of PAK2 association is highly variable among Nef proteins from different lentiviruses, prompting us to evaluate the conservation of this actin-remodeling/cofilin-deregulating mechanism. Based on the analysis of a total of 17 HIV-1, HIV-2, and SIV Nef proteins, we report here that inhibition of chemokine-induced actin remodeling as well as inactivation of cofilin are strongly conserved activities of lentiviral Nef proteins. Of note, even for Nef variants that display only marginal PAK2 association in vitro, these activities require the integrity of a PAK2 recruitment motif and the presence of endogenous PAK2. Thus, reduced in vitro affinity to PAK2 does not indicate limited functionality of Nef-PAK2 complexes in intact HIV-1 host cells. These results establish hijacking of PAK2 for deregulation of cofilin and inhibition of triggered actin remodeling as a highly conserved function of lentiviral Nef proteins, supporting the notion that PAK2 association may be critical for Nef's activity in vivo. PMID:20147394

  7. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP.

  8. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP

  9. Evaluasi Konsumen di "RM. Pak Kardi" Pemalang dengan Analisis Diskriminan

    Directory of Open Access Journals (Sweden)

    Tuis Susanto

    2017-02-01

    Full Text Available Penelitian ini bertujuan mengetahui apakah ada perbedaan yang siqnifikan dalam variabel dependen (Y yang meliputi konsumen sering Beli (YO, Cukup (Y I dan Jarang Beli (Y2, serta bertujuan mengetahui perilaku konsumen yang benar-benar berbeda, perbedaan dalam  arti perilaku   mereka  sering  membeli, cukup dan jarang membeli. Metode analisis  yang digunakan adalah dengan Wilk's Lambda, Pairwise, F test. Canonical  corellation,   untuk  mencari ada dan tidak perbedaan antar group variabel dependen dan  menginterpretasikan     berdasarkan function at group centroid untuk mengetahui variabel independen mana yang   menjadi  faktor diskriminannya. Hasil yang didapat dari penelitian ini adalah  bahwa  variabel  menu  merupakan faktor pembeda (diskriminan.  Artinya konsumen grup (sering  beli,  cukup  dan jarang   beli  tidak terpengaruh   dengan  usia,  harga,  pendapatan,   dan pelayanan  yang  diberikan oleh RM. Pak Kardi ditunjukkan dengan tanda (+ padafunction I. Jadi  konsumen  yang  membeli  di RM. Pak  Kardi adalah  mereka  yang  benar-benar   menyukai menu  (masakan  khas kepiting RM. Pak Kardi.  Jarak antara  grup  Sering  Beli  dengan grup  Jarang   Beli  adalah yang  terbesar, yakni 7,350. Sedangkan jarak   terkecil   adalah antara grup  Cukup  dengan  grup  Jarang   Beli  (0.522. Dengan demikian dapat  dikatakan   bahwa  Konsumen   di grup  Sering  Beli paling berbeda  selera  Menu  masakannya. Sebaliknya   Menu yang  disukai  oleh  konsumen di RM. Pak Kardi yang termasuk konsumen Cukup mempunyai perbedaan yang  kecil dengan  mereka  yang jarang membeli

  10. IAEA OSART/EXPERT follow-up review mission completes assessment of actions taken by Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    2004-01-01

    On 28 February 2004 the IAEA completed its follow-up review mission to assess the actions taken by Paks nuclear power plant (NPP) in response to the Agency's recommendations and suggestions made during the 2001 Operational Safety Review Team (OSART) mission and the 2003 Expert mission that investigated the fuel cleaning incident at the Paks NPP in Hungary. The mission was requested by the Hungarian Government to provide an independent assessment of the actions taken by Paks NPP. The IAEA team determined that the actions taken by Paks have resulted in tangible progress and concluded that all issues were either fully resolved or progressing satisfactorily. In a press conference, the team's conclusions in five areas were highlighted: management, Regulatory Oversight/Interface, operations and maintenance, including operating experience, radiation protection, emergency planning and preparedness, and transparency

  11. Pak pakub porgandi- ja peedipulbrit / Silja Lättemäe

    Index Scriptorium Estoniae

    Lättemäe, Silja, 1952-

    2006-01-01

    Harjumaa Kadarbiku köögiviljakasvatustalu peremees Ants Pak hakkab porgandi- ja peedimahla tootmisjääke pulbriks kuivatama. Uudistoodet on plaanis turustada kamalisandina ja ekspordiks arengumaadesse

  12. RIT1 controls actin dynamics via complex formation with RAC1/CDC42 and PAK1.

    Directory of Open Access Journals (Sweden)

    Uta Meyer Zum Büschenfelde

    2018-05-01

    Full Text Available RIT1 belongs to the RAS family of small GTPases. Germline and somatic RIT1 mutations have been identified in Noonan syndrome (NS and cancer, respectively. By using heterologous expression systems and purified recombinant proteins, we identified the p21-activated kinase 1 (PAK1 as novel direct effector of RIT1. We found RIT1 also to directly interact with the RHO GTPases CDC42 and RAC1, both of which are crucial regulators of actin dynamics upstream of PAK1. These interactions are independent of the guanine nucleotide bound to RIT1. Disease-causing RIT1 mutations enhance protein-protein interaction between RIT1 and PAK1, CDC42 or RAC1 and uncouple complex formation from serum and growth factors. We show that the RIT1-PAK1 complex regulates cytoskeletal rearrangements as expression of wild-type RIT1 and its mutant forms resulted in dissolution of stress fibers and reduction of mature paxillin-containing focal adhesions in COS7 cells. This effect was prevented by co-expression of RIT1 with dominant-negative CDC42 or RAC1 and kinase-dead PAK1. By using a transwell migration assay, we show that RIT1 wildtype and the disease-associated variants enhance cell motility. Our work demonstrates a new function for RIT1 in controlling actin dynamics via acting in a signaling module containing PAK1 and RAC1/CDC42, and highlights defects in cell adhesion and migration as possible disease mechanism underlying NS.

  13. RIT1 controls actin dynamics via complex formation with RAC1/CDC42 and PAK1.

    Science.gov (United States)

    Meyer Zum Büschenfelde, Uta; Brandenstein, Laura Isabel; von Elsner, Leonie; Flato, Kristina; Holling, Tess; Zenker, Martin; Rosenberger, Georg; Kutsche, Kerstin

    2018-05-01

    RIT1 belongs to the RAS family of small GTPases. Germline and somatic RIT1 mutations have been identified in Noonan syndrome (NS) and cancer, respectively. By using heterologous expression systems and purified recombinant proteins, we identified the p21-activated kinase 1 (PAK1) as novel direct effector of RIT1. We found RIT1 also to directly interact with the RHO GTPases CDC42 and RAC1, both of which are crucial regulators of actin dynamics upstream of PAK1. These interactions are independent of the guanine nucleotide bound to RIT1. Disease-causing RIT1 mutations enhance protein-protein interaction between RIT1 and PAK1, CDC42 or RAC1 and uncouple complex formation from serum and growth factors. We show that the RIT1-PAK1 complex regulates cytoskeletal rearrangements as expression of wild-type RIT1 and its mutant forms resulted in dissolution of stress fibers and reduction of mature paxillin-containing focal adhesions in COS7 cells. This effect was prevented by co-expression of RIT1 with dominant-negative CDC42 or RAC1 and kinase-dead PAK1. By using a transwell migration assay, we show that RIT1 wildtype and the disease-associated variants enhance cell motility. Our work demonstrates a new function for RIT1 in controlling actin dynamics via acting in a signaling module containing PAK1 and RAC1/CDC42, and highlights defects in cell adhesion and migration as possible disease mechanism underlying NS.

  14. Ex-core fuel damage event at paks causes, consequences and lessons learned

    International Nuclear Information System (INIS)

    Bajsz, J.; Gado, J.

    2004-01-01

    On April 10, 2003 Paks NPP experienced a loss of decay-heat removal to 30 irradiated fuel assemblies undergoing a cleaning process in a fuel service pit near the unit 2 spent fuel pool. Following chemical cleaning of high decay-heat fuel, a delay in removing the cleaning vessel's lid left the cleaning system in such a condition that did not provide adequate cooling to the fuel. After several hours of the fuel being under-cooled, a steam bubble developed in the vessel, essentially uncovering the fuel. When the lid of the vessel was removed, the sudden introduction of cool water thermally shocked the fuel causing significant structural damage and a release of fission product gases to the reactor building. The paper will discuss the causes of the event as well as the contributing factors to it. Detailed information will be given about the planning and preparation of the recovery actions. The in-depth analyses of the consequences and lessons learned complete the lecture. (author)

  15. Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

    1999-05-15

    The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

  16. How to ensure the safety of extended operations: Practice and experience of Paks NPP

    International Nuclear Information System (INIS)

    Kovacs, J.

    2005-01-01

    The Paks Nuclear Power Plant strategy is to extend the operational lifetime of the plant and renew the operational license for 20 years over the designed and licensed lifetime. In the paper the preconditions of long-term operation are discussed and the basic findings and experience of the license renewal works are also presented. The further plans fo NPP Paks for ensuring safe operation in long-term are discussed. (author)

  17. Rho Kinase (ROCK) collaborates with Pak to Regulate Actin Polymerization and Contraction in Airway Smooth Muscle.

    Science.gov (United States)

    Zhang, Wenwu; Bhetwal, Bhupal P; Gunst, Susan J

    2018-05-10

    The mechanisms by which Rho kinase (ROCK) regulates airway smooth muscle contraction were determined in tracheal smooth muscle tissues. ROCK may mediate smooth muscle contraction by inhibiting myosin regulatory light chain (RLC) phosphatase. ROCK can also regulate F-actin dynamics during cell migration, and actin polymerization is critical for airway smooth muscle contraction. Our results show that ROCK does not regulate airway smooth muscle contraction by inhibiting myosin RLC phosphatase or by stimulating myosin RLC phosphorylation. We find that ROCK regulates airway smooth muscle contraction by activating the serine-threonine kinase Pak, which mediates the activation of Cdc42 and Neuronal-Wiskott-Aldrich Syndrome protein (N-WASp). N-WASP transmits signals from cdc42 to the Arp2/3 complex for the nucleation of actin filaments. These results demonstrate a novel molecular function for ROCK in the regulation of Pak and cdc42 activation that is critical for the processes of actin polymerization and contractility in airway smooth muscle. Rho kinase (ROCK), a RhoA GTPase effector, can regulate the contraction of airway and other smooth muscle tissues. In some tissues, ROCK can inhibit myosin regulatory light chain (RLC) phosphatase, which increases the phosphorylation of myosin RLC and promotes smooth muscle contraction. ROCK can also regulate cell motility and migration by affecting F-actin dynamics. Actin polymerization is stimulated by contractile agonists in airway smooth muscle tissues and is required for contractile tension development in addition to myosin RLC phosphorylation. We investigated the mechanisms by which ROCK regulates the contractility of tracheal smooth muscle tissues by expressing a kinase inactive mutant of ROCK, ROCK-K121G, in the tissues or by treating them with the ROCK inhibitor, H-1152P. Our results show no role for ROCK in the regulation of non-muscle or smooth muscle myosin RLC phosphorylation during contractile stimulation in this tissue

  18. Anti-cancer effect of novel PAK1 inhibitor via induction of PUMA-mediated cell death and p21-mediated cell cycle arrest.

    Science.gov (United States)

    Woo, Tae-Gyun; Yoon, Min-Ho; Hong, Shin-Deok; Choi, Jiyun; Ha, Nam-Chul; Sun, Hokeun; Park, Bum-Joon

    2017-04-04

    Hyper-activation of PAK1 (p21-activated kinase 1) is frequently observed in human cancer and speculated as a target of novel anti-tumor drug. In previous, we also showed that PAK1 is highly activated in the Smad4-deficient condition and suppresses PUMA (p53 upregulated modulator of apoptosis) through direct binding and phosphorylation. On the basis of this result, we have tried to find novel PAK1-PUMA binding inhibitors. Through ELISA-based blind chemical library screening, we isolated single compound, IPP-14 (IPP; Inhibitor of PAK1-PUMA), which selectively blocks the PAK1-PUMA binding and also suppresses cell proliferation via PUMA-dependent manner. Indeed, in PUMA-deficient cells, this chemical did not show anti-proliferating effect. This chemical possessed very strong PAK1 inhibition activity that it suppressed BAD (Bcl-2-asoociated death promoter) phosphorylation and meta-phase arrest via Aurora kinase inactivation in lower concentration than that of previous PAK1 kinase, FRAX486 and AG879. Moreover, our chemical obviously induced p21/WAF1/CIP1 (Cyclin-dependent kinase inhibitor 1A) expression by releasing from Bcl-2 (B-cell lymphoma-2) and by inhibition of AKT-mediated p21 suppression. Considering our result, IPP-14 and its derivatives would be possible candidates for PAK1 and p21 induction targeted anti-cancer drug.

  19. Statistical analysis of the vibration loading of the reactor internals and fuel assemblies of reactor units type WWER-440 from deferent projects

    International Nuclear Information System (INIS)

    Ovcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.; Anikeev, J.; Pljush, A.

    2006-01-01

    In this paper the following items have been presented: 1) Vibration noise instrument channels; 2) Vibration loading characteristics of control assemblies, internals and design peculiarities of internals of WWER-440 deferent projects; 3) Coolant flow rate through the reactor, reactor core, fuel assemblies and control assemblies for different projects WWER-440 and 4) Noise measurements of coolant speed per channel. The change of auto power spectrum density of absolute displacement detector signal for the last 12 years of SUS monitoring of the Kola NPP unit 2; the coherence functions groups between two SPND of the same level for the Kola NPP unit 1; the measured coolant flow rate at Paks NPP and the auto power spectrum density group of SPND signals from 11 neutron measuring channels of the Kola NPP unit 1 are given. The main factors of vibration loading of internals and fuel assemblies for Kola NPP units 1-4, Bohunice NPP units 1 and 2 and Novovoronezh NPP units 3 and 4 are also discussed

  20. Increased expression of microRNA-221 inhibits PAK1 in endothelial progenitor cells and impairs its function via c-Raf/MEK/ERK pathway

    International Nuclear Information System (INIS)

    Zhang, Xiaoping; Mao, Haian; Chen, Jin-yuan; Wen, Shengjun; Li, Dan; Ye, Meng; Lv, Zhongwei

    2013-01-01

    Highlights: ► MicroRNA-221 is upregulated in the endothelial progenitor cells of atherosclerosis patients. ► PAK1 is a direct target of microRNA-221. ► MicroRNA-221 inhibits EPCs proliferation through c-Raf/MEK/ERK pathway. -- Abstract: Coronary artery disease (CAD) is associated with high mortality and occurs via endothelial injury. Endothelial progenitor cells (EPCs) restore the integrity of the endothelium and protect it from atherosclerosis. In this study, we compared the expression of microRNAs (miRNAs) in EPCs in atherosclerosis patients and normal controls. We found that miR-221 expression was significantly up-regulated in patients compared with controls. We predicted and identified p21/Cdc42/Rac1-activated kinase 1 (PAK1) as a novel target of miR-221 in EPCs. We also demonstrated that miR-221 targeted a putative binding site in the 3′UTR of PAK1, and absence of this site was inversely associated with miR-221 expression in EPCs. We confirmed this relationship using a luciferase reporter assay. Furthermore, overexpression of miR-221 in EPCs significantly decreased EPC proliferation, in accordance with the inhibitory effects induced by decreased PAK1. Overall, these findings demonstrate that miR-221 affects the MEK/ERK pathway by targeting PAK1 to inhibit the proliferation of EPCs

  1. Integrated solidity test measurement of the airtight compartment system at the Paks nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Osztheimer, M.; Taubner, R.; Techy, Zs. (Villamosenergiaipari Kutato Intezet, Budapest (Hungary))

    1983-01-01

    A brief report on the purpose of the integrated solidity test measurements of the airtight compartment system of the Paks nuclear power plant and on the applied measuring principles is given. The measuring system and the selected measuring methods are evaluated. The characteristic features of the airtight system of the Paks nuclear power plant's 1st block and their effects on the measurement are mentioned.

  2. Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

    Science.gov (United States)

    Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

    2010-10-01

    The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

  3. Introducing advanced ISI requirements at Paks NPP for supporting the LTO

    International Nuclear Information System (INIS)

    Trampus, P.; Ratkai, S.

    2012-01-01

    The four VVER-440 model 213 units in operation at Paks NPP, Hungary, are facing to approach their licensed term of operation, which is 30 years. To extend the safe operation of the units beyond the original licensed term by additional 20 years belongs to the highest priorities of the owners/operator of MVM Paks NPP. According to the nuclear legislation, a formal license renewal application for the extended period has to be submitted to the Hungarian Atomic Energy Authority. A significant feature of the license renewal process is the demonstration of the effectiveness of the currently applied ageing management program. ISI is an essential part of the ageing management program thus the adequate ISI techniques and the tailor made requirements have to be incorporated in it. To cope with the expectations originating from the LTO at Paks NPP, it was decided to replace the original Soviet based ISI system by the widely applied ASME BPVC Section XI requirements. Additionally, in 2011 a new nuclear regulation was issued in Hungary, in which the ISI requirements have also been changed. This paper intends to present the entire structure of the new Hungarian regulation related to the ISI but mainly focusing on the deviation to the ASME Section XI with the perspective of the licence renewal. (author)

  4. Numerical analyses of an ex-core fuel incident: Results of the OECD-IAEA Paks Fuel Project

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z., E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Aszodi, A. [BME NTI Budapest (Hungary); Barnak, M. [IVS, Trnava (Slovakia); Boros, I. [BME NTI Budapest (Hungary); Fogel, M. [VUJE, Trnava (Slovakia); Guillard, V. [IRSN, Cadarache (France); Gyori, Cs. [ITU, EU, Karlsruhe (Germany); Hegyi, G. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Horvath, G.L. [VEIKI, Budapest (Hungary); Nagy, I. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Junninen, P. [VTT, Espoo (Finland); Kobzar, V. [KI, Moscow (Russian Federation); Legradi, G. [BME NTI Budapest (Hungary); Molnar, A. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Pietarinen, K. [VTT, Espoo (Finland); Perneczky, L. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Makihara, Y. [ATMEA, Paris (France); Matejovic, P. [IVS, Trnava (Slovakia); Perez-Fero, E.; Slonszki, E. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary)

    2010-03-15

    The OECD-IAEA Paks Fuel Project was developed to support the understanding of fuel behaviour in accident conditions on the basis of analyses of the Paks-2 incident. Numerical simulation of the most relevant aspects of the event and comparison of the calculation results with the available data from the incident was carried out between 2006 and 2007. A database was compiled to provide input for the code calculations. The activities covered the following three areas: (a) Thermal hydraulic calculations described the cooling conditions possibly established during the incident. (b) Simulation of fuel behaviour described the oxidation and degradation mechanisms of the fuel assemblies. (c) The release of fission products from the failed fuel rods was estimated and compared to available measured data. The applied used codes captured the most important events of the Paks-2 incident and the calculated results improved the understanding of the causes and mechanisms of fuel failure. The numerical analyses showed that the by-pass flow leading to insufficient cooling amounted to 75-90% of the inlet flow rate, the maximum temperature in the tank was between 1200 and 1400 deg. C, the degree of zirconium oxidation reached 4-12% and the mass of produced hydrogen was between 3 and 13 kg.

  5. Synapses of Amphids Defective (SAD-A) Kinase Promotes Glucose-stimulated Insulin Secretion through Activation of p21-activated Kinase (PAK1) in Pancreatic β-Cells*

    Science.gov (United States)

    Nie, Jia; Sun, Chao; Faruque, Omar; Ye, Guangming; Li, Jia; Liang, Qiangrong; Chang, Zhijie; Yang, Wannian; Han, Xiao; Shi, Yuguang

    2012-01-01

    The p21-activated kinase-1 (PAK1) is implicated in regulation of insulin exocytosis as an effector of Rho GTPases. PAK1 is activated by the onset of glucose-stimulated insulin secretion (GSIS) through phosphorylation of Thr-423, a major activation site by Cdc42 and Rac1. However, the kinase(s) that phosphorylates PAK1 at Thr-423 in islet β-cells remains elusive. The present studies identified SAD-A (synapses of amphids defective), a member of AMP-activated protein kinase-related kinases exclusively expressed in brain and pancreas, as a key regulator of GSIS through activation of PAK1. We show that SAD-A directly binds to PAK1 through its kinase domain. The interaction is mediated by the p21-binding domain (PBD) of PAK1 and requires both kinases in an active conformation. The binding leads to direct phosphorylation of PAK1 at Thr-423 by SAD-A, triggering the onset of GSIS from islet β-cells. Consequently, ablation of PAK1 kinase activity or depletion of PAK1 expression completely abolishes the potentiating effect of SAD-A on GSIS. Consistent with its role in regulating GSIS, overexpression of SAD-A in MIN6 islet β-cells significantly stimulated cytoskeletal remodeling, which is required for insulin exocytosis. Together, the present studies identified a critical role of SAD-A in the activation of PAK1 during the onset of insulin exocytosis. PMID:22669945

  6. Synapses of amphids defective (SAD-A) kinase promotes glucose-stimulated insulin secretion through activation of p21-activated kinase (PAK1) in pancreatic β-Cells.

    Science.gov (United States)

    Nie, Jia; Sun, Chao; Faruque, Omar; Ye, Guangming; Li, Jia; Liang, Qiangrong; Chang, Zhijie; Yang, Wannian; Han, Xiao; Shi, Yuguang

    2012-07-27

    The p21-activated kinase-1 (PAK1) is implicated in regulation of insulin exocytosis as an effector of Rho GTPases. PAK1 is activated by the onset of glucose-stimulated insulin secretion (GSIS) through phosphorylation of Thr-423, a major activation site by Cdc42 and Rac1. However, the kinase(s) that phosphorylates PAK1 at Thr-423 in islet β-cells remains elusive. The present studies identified SAD-A (synapses of amphids defective), a member of AMP-activated protein kinase-related kinases exclusively expressed in brain and pancreas, as a key regulator of GSIS through activation of PAK1. We show that SAD-A directly binds to PAK1 through its kinase domain. The interaction is mediated by the p21-binding domain (PBD) of PAK1 and requires both kinases in an active conformation. The binding leads to direct phosphorylation of PAK1 at Thr-423 by SAD-A, triggering the onset of GSIS from islet β-cells. Consequently, ablation of PAK1 kinase activity or depletion of PAK1 expression completely abolishes the potentiating effect of SAD-A on GSIS. Consistent with its role in regulating GSIS, overexpression of SAD-A in MIN6 islet β-cells significantly stimulated cytoskeletal remodeling, which is required for insulin exocytosis. Together, the present studies identified a critical role of SAD-A in the activation of PAK1 during the onset of insulin exocytosis.

  7. COMPARISON OF PAI AND PAK: AN OVERVIEW OF VALUES OF MULTICULTURAL EDUCATION

    Directory of Open Access Journals (Sweden)

    Ali Murfi

    2015-06-01

    Full Text Available This research to reveal comparative Islamic Education (PAI with Christian Education (PAK through a textbook’s lesson in terms of content values of multicultural education. The comparative’s analysis includes three aspects, differences, similarities, and common platform. The results showed that substance of values of multicultural education contained in the textbooks have much in similarities which eventually became common platform both than the differences that exist, so that PAI and PAK should move bind themselves to each other in one joint effort to raise the noble values of multicultural, where both scientific traditions stand firm through efforts integration and comprehension charge of teaching materials.

  8. SparsePak: A Formatted Fiber Field Unit for the WIYN Telescope Bench Spectrograph. I. Design, Construction, and Calibration

    NARCIS (Netherlands)

    Bershady, Matthew A.; Andersen, David R.; Harker, Justin; Ramsey, Larry W.; Verheijen, Marc A. W.

    2004-01-01

    We describe the design and construction of a formatted fiber field unit, SparsePak, and characterize its optical and astrometric performance. This array is optimized for spectroscopy of low surface brightness extended sources in the visible and near-infrared. SparsePak contains 82, 4.7" fibers

  9. Stereoselective uptake and distribution of the chiral neonicotinoid insecticide, Paichongding, in Chinese pak choi (Brassica campestris ssp. chinenesis)

    International Nuclear Information System (INIS)

    Wang, Haiyan; Yang, Zhen; Liu, Ruyang; Fu, Qiuguo; Zhang, Sufen; Cai, Zhiqiang; Li, Juying; Zhao, Xiaojun; Ye, Qingfu; Wang, Wei; Li, Zhong

    2013-01-01

    Highlights: • Absorption of foliar applied Paichongding by pak choi was not stereoselective. • Foliar uptake and downward transport of Paichongding were both found in pak choi. • Enantioselective and epimer-selective root uptake were observed for Paichongding. • Foliage/root uptake showed diastereoselective transport of Paichongding epimers. • The SR and RS are more easily taken up by roots and accumulated in edible parts. -- Abstract: Neonicotinoid chiral insecticidal Paichongding is a promising substitute for the widely used imidacloprid. Four stereoisomers of Paichongding, 5R,7R, 5S,7S, 5S,7R and 5R,7S, were employed in both foliage and roots of Chinese pak choi to investigate their stereoselective uptake and distribution in pak choi. Results showed that after foliar application, no stereoselective absorption into pak-choi plants was observed among the enantiomers. Total absorptions were 35.40% of the applied amount for 5R,7R, 36.66% for 5S,7S, 36.80% for 5S,7R and 38.20% for 5R,7S at 96 HAT. The translocation of the four absorbed stereoisomers within pak choi occurred both acropetally and basipetally and the transport of 14 C from enantiomers 5R,7R and 5S,7S were significantly higher than for 5R,7S and 5S,7R. Significant stereoselective translocation inside plants was observed between Paichongding epimers. Total root uptake reached 16.49–19.85% for 5R,7R and 5S,7S, and 24.57–28.82% for 5S,7R and 5R,7S at 144 HAT. Both enantioselective and diastereoselective root uptake into pak-choi occurred between the four stereoisomers. The 5R,7S and 5S,7R enantiomers were more readily uptaken by the roots than 5R,7R and 5S,7S and accumulated in the edible leaves. These results will help to develop an understanding of Paichongding using only the target-active enantiomer of pesticides

  10. Stereoselective uptake and distribution of the chiral neonicotinoid insecticide, Paichongding, in Chinese pak choi (Brassica campestris ssp. chinenesis)

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Haiyan; Yang, Zhen; Liu, Ruyang; Fu, Qiuguo; Zhang, Sufen; Cai, Zhiqiang; Li, Juying; Zhao, Xiaojun [Institute of Nuclear Agricultural Sciences, Key Laboratory of Nuclear Agricultural Sciences of Ministry of Agriculture and Zhejiang Province, Zhejiang University, Hangzhou 310029 (China); Ye, Qingfu, E-mail: qfye@zju.edu.cn [Institute of Nuclear Agricultural Sciences, Key Laboratory of Nuclear Agricultural Sciences of Ministry of Agriculture and Zhejiang Province, Zhejiang University, Hangzhou 310029 (China); Wang, Wei [Institute of Nuclear Agricultural Sciences, Key Laboratory of Nuclear Agricultural Sciences of Ministry of Agriculture and Zhejiang Province, Zhejiang University, Hangzhou 310029 (China); Li, Zhong, E-mail: lizhong@ecust.edu.cn [School of Pharmacy, East China University of Science and Technology, 130 Meilong Road, Shanghai 200237 (China)

    2013-11-15

    Highlights: • Absorption of foliar applied Paichongding by pak choi was not stereoselective. • Foliar uptake and downward transport of Paichongding were both found in pak choi. • Enantioselective and epimer-selective root uptake were observed for Paichongding. • Foliage/root uptake showed diastereoselective transport of Paichongding epimers. • The SR and RS are more easily taken up by roots and accumulated in edible parts. -- Abstract: Neonicotinoid chiral insecticidal Paichongding is a promising substitute for the widely used imidacloprid. Four stereoisomers of Paichongding, 5R,7R, 5S,7S, 5S,7R and 5R,7S, were employed in both foliage and roots of Chinese pak choi to investigate their stereoselective uptake and distribution in pak choi. Results showed that after foliar application, no stereoselective absorption into pak-choi plants was observed among the enantiomers. Total absorptions were 35.40% of the applied amount for 5R,7R, 36.66% for 5S,7S, 36.80% for 5S,7R and 38.20% for 5R,7S at 96 HAT. The translocation of the four absorbed stereoisomers within pak choi occurred both acropetally and basipetally and the transport of {sup 14}C from enantiomers 5R,7R and 5S,7S were significantly higher than for 5R,7S and 5S,7R. Significant stereoselective translocation inside plants was observed between Paichongding epimers. Total root uptake reached 16.49–19.85% for 5R,7R and 5S,7S, and 24.57–28.82% for 5S,7R and 5R,7S at 144 HAT. Both enantioselective and diastereoselective root uptake into pak-choi occurred between the four stereoisomers. The 5R,7S and 5S,7R enantiomers were more readily uptaken by the roots than 5R,7R and 5S,7S and accumulated in the edible leaves. These results will help to develop an understanding of Paichongding using only the target-active enantiomer of pesticides.

  11. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4D. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on seismic margin assessment and earthquake experience based methods for WWER-440/213 type NPPs; structural analysis and site inspection for site requalification; structural response of Paks NPP reactor building; analysis and testing of model worm type tanks on shaking table; vibration test of a worm tank model; evaluation of potential hazard for operating WWER control rods under seismic excitation

  12. Energy policy, economic and engineering issues of the extension of Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Aszodi, Attila; Boros, Ildiko; Kovacs, Arnold

    2014-01-01

    The four operating blocks of the Paks Nuclear Power Plant are of Russian design. They entered into operation three decades ago, between 1982 and 1987. In 2013 they produced 15 TWh out of the 42 TWh energy consumption of Hungary, that is they produced 36% of the energy demand. In the beginning of 2014 the Hungarian and the Russian governments signed the agreement on the extension of Paks site with building two new blocks, producing 1200 MW each. The paper summarizes the energy policy, engineering, safety and economic aspects of the extension. (TRA)

  13. Composition and activity variations in bulk gas of drum waste packages of Paks NPP

    International Nuclear Information System (INIS)

    Molnar, M.; Palcsu, L.; Svingor, E.; Szanto, Zs.; Futo, I.; Ormai, P.

    2001-01-01

    To obtain reliable estimates of the quantities and rates of the gas production a series of measurements was carried out in drum waste packages generated and temporarily stored at the site of Paks Nuclear Power Plant (Paks NPP). Ten drum waste packages were equipped with sampling valves for repeated sampling. Nine times between 04/02/2000 and 19/07/2001 qualitative gas component analyses of bulk gases of drums were executed. Gas samples were delivered to the laboratory of the ATOMKI for tritium and radiocarbon content measurements.(author)

  14. Induced Production of 1-Methoxy-indol-3-ylmethyl Glucosinolate by Jasmonic Acid and Methyl Jasmonate in Sprouts and Leaves of Pak Choi (Brassica rapa ssp. chinensis

    Directory of Open Access Journals (Sweden)

    Hansruedi Glatt

    2013-07-01

    Full Text Available Pak choi plants (Brassica rapa ssp. chinensis were treated with different signaling molecules methyl jasmonate, jasmonic acid, linolenic acid, and methyl salicylate and were analyzed for specific changes in their glucosinolate profile. Glucosinolate levels were quantified using HPLC-DAD-UV, with focus on induction of indole glucosinolates and special emphasis on 1-methoxy-indol-3-ylmethyl glucosinolate. Furthermore, the effects of the different signaling molecules on indole glucosinolate accumulation were analyzed on the level of gene expression using semi-quantitative realtime RT-PCR of selected genes. The treatments with signaling molecules were performed on sprouts and mature leaves to determine ontogenetic differences in glucosinolate accumulation and related gene expression. The highest increase of indole glucosinolate levels, with considerable enhancement of the 1-methoxy-indol-3-ylmethyl glucosinolate content, was achieved with treatments of sprouts and mature leaves with methyl jasmonate and jasmonic acid. This increase was accompanied by increased expression of genes putatively involved in the indole glucosinolate biosynthetic pathway. The high levels of indole glucosinolates enabled the plant to preferentially produce the respective breakdown products after tissue damage. Thus, pak choi plants treated with methyl jasmonate or jasmonic acid, are a valuable tool to analyze the specific protection functions of 1-methoxy-indole-3-carbinole in the plants defense strategy in the future.

  15. Slit stimulation recruits Dock and Pak to the roundabout receptor and increases Rac activity to regulate axon repulsion at the CNS midline.

    Science.gov (United States)

    Fan, Xueping; Labrador, Juan Pablo; Hing, Huey; Bashaw, Greg J

    2003-09-25

    Drosophila Roundabout (Robo) is the founding member of a conserved family of repulsive axon guidance receptors that respond to secreted Slit proteins. Here we present evidence that the SH3-SH2 adaptor protein Dreadlocks (Dock), the p21-activated serine-threonine kinase (Pak), and the Rac1/Rac2/Mtl small GTPases can function during Robo repulsion. Loss-of-function and genetic interaction experiments suggest that limiting the function of Dock, Pak, or Rac partially disrupts Robo repulsion. In addition, Dock can directly bind to Robo's cytoplasmic domain, and the association of Dock and Robo is enhanced by stimulation with Slit. Furthermore, Slit stimulation can recruit a complex of Dock and Pak to the Robo receptor and trigger an increase in Rac1 activity. These results provide a direct physical link between the Robo receptor and an important cytoskeletal regulatory protein complex and suggest that Rac can function in both attractive and repulsive axon guidance.

  16. Dbo/Henji Modulates Synaptic dPAK to Gate Glutamate Receptor Abundance and Postsynaptic Response.

    Science.gov (United States)

    Wang, Manyu; Chen, Pei-Yi; Wang, Chien-Hsiang; Lai, Tzu-Ting; Tsai, Pei-I; Cheng, Ying-Ju; Kao, Hsiu-Hua; Chien, Cheng-Ting

    2016-10-01

    In response to environmental and physiological changes, the synapse manifests plasticity while simultaneously maintains homeostasis. Here, we analyzed mutant synapses of henji, also known as dbo, at the Drosophila neuromuscular junction (NMJ). In henji mutants, NMJ growth is defective with appearance of satellite boutons. Transmission electron microscopy analysis indicates that the synaptic membrane region is expanded. The postsynaptic density (PSD) houses glutamate receptors GluRIIA and GluRIIB, which have distinct transmission properties. In henji mutants, GluRIIA abundance is upregulated but that of GluRIIB is not. Electrophysiological results also support a GluR compositional shift towards a higher IIA/IIB ratio at henji NMJs. Strikingly, dPAK, a positive regulator for GluRIIA synaptic localization, accumulates at the henji PSD. Reducing the dpak gene dosage suppresses satellite boutons and GluRIIA accumulation at henji NMJs. In addition, dPAK associated with Henji through the Kelch repeats which is the domain essential for Henji localization and function at postsynapses. We propose that Henji acts at postsynapses to restrict both presynaptic bouton growth and postsynaptic GluRIIA abundance by modulating dPAK.

  17. Three-dimensional reactor model for the Paks NPP full-scope simulator

    International Nuclear Information System (INIS)

    Gyori, C.; Hegyi, G.; Hozer, Z.; Kereszturi, A.; Maraczy, C.

    1993-01-01

    The reactor model includes thermohydraulic and neutron-physical components. The thermohydraulic model is based on the SMABRE code developed at the Technical Research Centre of Finland for the analysis of loss-of-coolant transients in PWRs. The fuel rod model will be replaced by a new software module providing a comprehensive description of the behavior of fuel rods during reactor transients and hypothetical accidents. The calculation is performed in four individual models: fuel rod temperature model, fuel rod internal pressure model, fuel rod deformation model and fuel rod failure model. In the neutron-physical model the core is calculated with nodes for all of the 349 fuel assemblies, and each assembly is calculated in ten layers. (Z.S.) 1 fig., 5 refs

  18. Migration of itx (Isopropyl Thioxantone from Tetra Pak Bricks into Food

    Directory of Open Access Journals (Sweden)

    Sonja Jamnicki

    2010-01-01

    Full Text Available At the beginning of September 2005, itx, a photoinitiator used in uv cured ink, has been identified to have migrated from packaging to food products. Tetra Pak has identified the source of migration to be uv cured offset printing ink.The presence of itx in food packed in Tetra Pak bricks is the result of the contamination of the inner polyethylene layer of the box walls. itx can either migrate through the packaging material or it can reach the food by contact, for example, as a result of the print set-off phenomenon. Most likely, the transfer of itx was due to the physical contact between the printed outer layer with the inner layer of the packaging, whereby the ink or ink substance transfers from the print to the reverse of the adjacent sheet.Tetra Pak has committed itself to move away from this technology immediately and to use alternative printing technologies to ensure that there is no or minimal migration of itx or other substances from its packages.itx is still not on the eu’s negative list of banned substances in food nor does the World Health Organization (who categorize it as being detrimental to human health. After an investigation in the health risks of itx following the incident, the European Food Safety Authority (efsa concluded that the levels found in foods, “while undesirable, do not give cause for health concern.”

  19. Migration of ITX (Isopropyl Thioxantone from Tetra Pak Bricks into Food

    Directory of Open Access Journals (Sweden)

    Tatjana Jamnicki

    2010-06-01

    Full Text Available At the beginning of September 2005, ITX, a photoinitiator used in uv cured ink, has been identified to have migrated from packaging to food products. Tetra Pak has identified the source of migration to be uv cured offset printing ink.The presence of ITX in food packed in Tetra Pak bricks is the result of the contamination of the inner polyethylene layer of the box walls. ITX can either migrate through the packaging material or it can reach the food by contact, for example, as a result of the print set-off phenomenon. Most likely, the transfer of ITX was due to the physical contact between the printed outer layer with the inner layer of the packaging, whereby the ink or ink substance transfers from the print to the reverse of the adjacent sheet.Tetra Pak has committed itself to move away from this technology immediately and to use alternative printing technologies to ensure that there is no or minimal migration of ITX or other substances from its packages.ITX is still not on the eu’s negative list of banned substances in food nor does the World Health Organization (WHO categorize it as being detrimental to human health. After an investigation in the health risks of ITX following the incident, the European Food Safety Authority (EFSA concluded that the levels found in foods, “while undesirable, do not give cause for health concern.”

  20. Stereoselective uptake and distribution of chiral neoniconoid insecticide paichongding in chinese pak choi (Brassica campestris ssp. Chinenesis)

    International Nuclear Information System (INIS)

    Wang Haiyan; Yang Zhen; Liu Ruyang; Fu Qiuguo; Zhang Sufen; Li Juying; Zhao Xiaojun; Ye Qingfu; Wang Wei; Li Zhong

    2014-01-01

    Paichongding, a neonicotinoid chiral insecticide containing two chiral centers, is a promising substitute for the widely used imidacloprid because it is effective against many imidacloprid-resistant insects. In this study, four optically-pure stereoisomers of Paichongding with 5R, 7R, 5S, 7S, 5S, 7R, and 5R, 7S were employed in both foliar and root of Chinese pak choi to investigate the stereoselective uptake and distribution of the insecticide in pak choi. The results showed, after foliar application, total absorption of individual "1"4C-Paichongding stereoisomers into pak-choi plants demonstrated no stereoselectivity between the enantiomers. The translocation of the four absorbed stereoisomers within pak choi occurred in both acropetal and basipetal directions and the transport of "1"4C from enantiomers 5R, 7R and 5S, 7S were significantly higher than enantiomers 5R, 7S and 5S, 7R. The statistically significant stereoselective translocation inside plants was observed between Paichongding epimers. Root treatment revealed that enantioselective and diastereoselective root uptake into pak-choi plants were both found between the four enantiomers. The enantiomers of 5R, 7S and 5S, 7R were more readily taken up by roots, and more readily accumulated in edible leaves than 5R, 7R and 5S, 7S. These results will help to develop an understanding of the proper application of Paichongding isomers in vegetables, and give useful information for food and environmental assessments of chiral pesticides. (authors)

  1. JENDL-3.3 thermal reactor benchmark test

    International Nuclear Information System (INIS)

    Akie, Hiroshi

    2001-01-01

    Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

  2. Discovery of 2-(4-Substituted-piperidin/piperazine-1-yl-N-(5-cyclopropyl-1H-pyrazol-3-yl-quinazoline-2,4-diamines as PAK4 Inhibitors with Potent A549 Cell Proliferation, Migration, and Invasion Inhibition Activity

    Directory of Open Access Journals (Sweden)

    Tianxiao Wu

    2018-02-01

    Full Text Available A series of novel 2,4-diaminoquinazoline derivatives were designed, synthesized, and evaluated as p21-activated kinase 4 (PAK4 inhibitors. All compounds showed significant inhibitory activity against PAK4 (half-maximal inhibitory concentration IC50 < 1 μM. Among them, compounds 8d and 9c demonstrated the most potent inhibitory activity against PAK4 (IC50 = 0.060 μM and 0.068 μM, respectively. Furthermore, we observed that compounds 8d and 9c displayed potent antiproliferative activity against the A549 cell line and inhibited cell cycle distribution, migration, and invasion of this cell line. In addition, molecular docking analysis was performed to predict the possible binding mode of compound 8d. This series of compounds has the potential for further development as PAK4 inhibitors for anticancer activity.

  3. External Events PSA for the Paks NPP

    International Nuclear Information System (INIS)

    Bareith, Attila; Karsa, Zoltan; Siklossy, Tamas; Vida, Zoltan

    2014-01-01

    Initially, probabilistic safety assessment of external events was limited to the analysis of earthquakes for the Paks Nuclear Power Plant in Hungary. The level 1 seismic PSA was completed in 2002 showing a significant contribution of seismic failures to core damage risk. Although other external events of natural origin had previously been screened out from detailed plant PSA mostly on the basis of event frequencies, a review of recent experience on extreme weather phenomena made during the periodic safety review of the plant led to the initiation of PSA for external events other than earthquakes in 2009. In the meantime, the accident of the Fukushima Dai-ichi Nuclear Power Plant confirmed further the importance of such an analysis. The external event PSA for the Paks plant followed the commonly known steps: selection and screening of external hazards, hazard assessment for screened-in external events, analysis of plant response and fragility, PSA model development, and risk quantification and interpretation of results. As a result of event selection and screening the following weather related external hazards were subject to detailed analysis: extreme wind, extreme rainfall (precipitation), extreme snow, extremely high and extremely low temperatures, lightning, frost and ice formation. The analysis proved to be a significant challenge due to scarcity of data, lack of knowledge, as well as limitations of existing PSA methodologies. This paper presents an overview of the external events PSA performed for the Paks NPP. Important methodological aspects are summarised. Key analysis findings and unresolved issues that need further elaboration are highlighted. Development of external events PSA for the Paks NPP was completed by the end of 2012. The analysis followed the commonly known steps: selection and screening of external hazards, hazard assessment for screened-in external events, analysis of plant response and fragility, PSA model development, and risk

  4. Assessment of Human Performance and Safety Culture at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Toth, Janos; Hadnagy, Lajos

    2002-01-01

    Evaluation of human performance and safety culture of the personnel at a Nuclear Power Plant is a very important element of the self assessment process. At the Paks NPP a systematic approach to this problem started in the early 90's. The first comprehensive analysis of the human performance of the personnel was performed by the Hungarian Research Institute for Electric Power (VEIKI). The analysis of human failures is also a part of the investigation and analysis of safety related reported events. This human performance analysis of events is carried out by the Laboratory of Psychology of the plant and a supporting organisation namely the Department of Ergonomics and Psychology of the Budapest University of Technical and Economical Sciences. The analysis of safety culture at the Paks NPP has been in the focus of attention since the implementation of the INSAG-4 document started world-wide. In 1993 an IAEA model project namely 'Strengthening Training for Operational Safety' was initiated with a sub-project called 'Enhancement of Safety Culture'. Within this project the first step was the initial assessment of the safety culture level at the Paks NPP. It was followed by some corrective actions and safety culture improvement programme. In 1999 the second assessment was performed in order to evaluate the progress as a result of the improvement programme. A few indicators reflecting the elements of safety culture were defined and compared. The assessment of the safety culture with a survey among the managers was performed in September 2000 and the results are being evaluated at the moment. The intention of the plant management is to repeat the assessment every 2-3 years and evaluate the trend of the indicator. (authors)

  5. Example of an application of systematic approach to training. Paks NPP Ltd

    International Nuclear Information System (INIS)

    Bajor, L.

    1998-01-01

    In order to show a practical example of how Hungarian Maintenance program works implementation of SAT based training of primary circuit field operator is described in detail as implemented at Paks NPP

  6. Pak2 Controls Acquisition of NKT Cell Fate by Regulating Expression of the Transcription Factors PLZF and Egr2

    Science.gov (United States)

    O’Hagan, Kyle L.; Zhao, Jie; Pryshchep, Olga; Wang, Chyung-Ru

    2015-01-01

    NKT cells constitute a small population of T cells developed in the thymus that produce large amounts of cytokines and chemokines in response to lipid Ags. Signaling through the Vα14-Jα18 TCR instructs commitment to the NKT cell lineage, but the precise signaling mechanisms that instruct their lineage choice are unclear. In this article, we report that the cytoskeletal remodeling protein, p21-activated kinase 2 (Pak2), was essential for NKT cell development. Loss of Pak2 in T cells reduced stage III NKT cells in the thymus and periphery. Among different NKT cell subsets, Pak2 was necessary for the generation and function of NKT1 and NKT2 cells, but not NKT17 cells. Mechanistically, expression of Egr2 and promyelocytic leukemia zinc finger (PLZF), two key transcription factors for acquiring the NKT cell fate, were markedly diminished in the absence of Pak2. Diminished expression of Egr2 and PLZF were not caused by aberrant TCR signaling, as determined using a Nur77-GFP reporter, but were likely due to impaired induction and maintenance of signaling lymphocyte activation molecule 6 expression, a TCR costimulatory receptor required for NKT cell development. These data suggest that Pak2 controls thymic NKT cell development by providing a signal that links Egr2 to induce PLZF, in part by regulating signaling lymphocyte activation molecule 6 expression. PMID:26519537

  7. Seismic safety of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Katona, T.

    1993-01-01

    An extensive program is underway at Paks NPP for evaluation of the seismic safety and for development of the necessary safety increasing measures. This program includes the following five measures: investigation of methods, regulations and techniques utilized for reassessment of seismic safety of operating NPPs and promoting safety; investigation of earthquake hazards; development of concepts for creating the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept, and preliminary evaluation of the seismic safety

  8. Controlled composting of waste wood contaminated with PAH; Untersuchungen zur gesteuerten Rotte von mit polyzyklischen aromatischen Kohlenwasserstoffen (PAK) kontaminiertem Altholz

    Energy Technology Data Exchange (ETDEWEB)

    Ulbricht, H.

    2002-07-01

    The author investigated the potential and limits of microbial pollutant degradation in PAH-polluted waste wood by composting. The conditions in which autochthonic micro-organisms are able to decomposite the PAH contained in wood by solid phase fermentation were investigated. The focus was on phenanthrene, anthracene and pyrene, all of which are used as protective materials (disinfestants) for wood. The results were verified on contaminated waste wood, including an analytical investigations of decomposition of PAH of the EPA catalogue. Boundary conditions for achieving high rates of PAH decomposition were investigated. [German] Generelles Ziel der Arbeit war die Untersuchung der Moeglichkeiten und Grenzen des mikrobiellen Schadstoffabbaus in PAK-belastetem Altholz durch Kompostierung und die Pruefung auf Anwendbarkeit der Erkenntnisse in technischen Verfahren. In der vorliegenden Arbeit wurde untersucht, unter welchen Bedingungen die autochthonen Mikroorganismen in der Lage sind, an das Holz gebundene PAK durch Feststofffermentation abzubauen. Als Schwerpunkt wurde zunaechst der Abbau der im zum Holzschutz verwendetem Teeroel vorkommenden PAK Phenanthren, Anthracen und Pyren untersucht. Eine Verifizierung der Ergebnisse erfolgte mit real kontaminiertem Altholz, dabei wurde der Abbau der PAK der EPA-Liste analytisch verfolgt. Es sollten geeignete Randbedingungen gefunden werden, um im Festphasensystem hohe Abbauraten der PAK zu erreichen. (orig.)

  9. Comparison of chromatography and Sep-pak methods for estimating the radiochemical purity of I-123 and I-131 labelled meta-iodobenzylguanidine (mIBG), synthesised in house

    International Nuclear Information System (INIS)

    Kumar, V.

    1998-01-01

    Full text: Radioiodine (I-123 or I-131) labelled mlBG has been prepared routinely in-house in a number of radiopharmacy laboratories. The radiochemical purity can be estimated by several methods. Available literature suggests that the results of chronatographic analysis are comparable with electrophoresis and high pressure liquid chromatography (HPLC) methods which are considered as gold-standard procedures. However, due to the cost involved with these equipments most of the radiopharmacy laboratories are not fortunate enough to have them. The present study compares the validity of reverse-phase Sep-pak cartridge method against chromatographic technique. We analysed twenty four preparations of mIBG by both Sep-pak and chromatography methods (20 batches of I-123 mD3G and 4 batches of I-131 mIBG). Chromatographic analysis, which takes >2hrs, was performed with Whatman No 1/ n-butanol: acetic acid: water (60:15:25 v/v) and the activity associated with the peaks for free iodine and I-123 mD3G were measured. Sep-pak cartridge method, which takes less than 10 min, was performed as follows: the cartridge was activated by injecting 5 mL ethanol (200% pure) followed by flushing with 5mL distilled water. A sample (0.1mL) of radioiodine labelled mD3G was applied to the column and was eluted with 5mL distilled water. Subsequently two aliquots of 5mL solution containing tetrahydrofuran: (0.1M) sodium dihydrogen phosphate (25:75v/v) were passed through and the activity in each elute was measured. Analysing the results by Student's paired t-test for I-123 mlBG using the Sep-pak method gave a mean + SD of 98.8+ 0.6 % which correlated well (r 2 = 0.780) with the results obtained by the chromatographic method 99.3+0.5% (p <0.05). The results obtained for free I-123 by the two methods were 1.09 + 0.56% and 0.6 + 0.5% (p <0.05) respectively. The parameters did not differ significantly when I-131, instead of I-123, was used to synthesise mIBG. The results clearly indicate that the Sep-pak

  10. Increased Rac1 activity and Pak1 overexpression are associated with lymphovascular invasion and lymph node metastasis of upper urinary tract cancer

    International Nuclear Information System (INIS)

    Kamai, Takao; Shirataki, Hiromichi; Nakanishi, Kimihiro; Furuya, Nobutaka; Kambara, Tsunehito; Abe, Hideyuki; Oyama, Tetsunari; Yoshida, Ken-Ichiro

    2010-01-01

    Lymphovascular invasion (LVI) and lymph node metastasis are conventional pathological factors associated with an unfavorable prognosis of urothelial carcinoma of the upper urinary tract (UC-UUT), but little is known about the molecular mechanisms underlying LVI and nodal metastasis in this disease. Rac1 small GTPase (Rac1) is essential for tumor metastasis. Activated GTP-bound Rac1 (Rac1 activity) plays a key role in activating downstream effectors known as Pak (21-activated kinase), which are key regulators of cytoskeletal remolding, cell motility, and cell proliferation, and thus have a role in both carcinogenesis and tumor invasion. We analyzed Rac1 activity and Pak1 protein expression in matched sets of tumor tissue, non-tumor tissue, and metastatic lymph node tissue obtained from the surgical specimens of 108 Japanese patients with UC-UUT. Rac1 activity and Pak1 protein levels were higher in tumor tissue and metastatic lymph node tissue than in non-tumor tissue (both P < 0.0001). A high level of Rac1 activity and Pak1 protein expression in the primary tumor was related to poor differentiation (P < 0.05), muscle invasion (P < 0.01), LVI (P < 0.0001), and lymph node metastasis (P < 0.0001). Kaplan-Meier survival analysis showed that an increase of Rac1 activity and Pak1 protein was associated with a shorter disease-free survival time (P < 0.01) and shorter overall survival (P < 0.001). Cox proportional hazards analysis revealed that high Rac1 activity, Pak1 protein expression and LVI were independent prognostic factors for shorter overall and disease-free survival times (P < 0.01) on univariate analysis, although only Pak1 and LVI had an influence (P < 0.05) according to multivariate analysis. These findings suggest that Rac1 activity and Pak1 are involved in LVI and lymph node metastasis of UC-UUT, and may be prognostic markers for this disease

  11. Results of a benchmark study for the seismic analysis and testing of WWER type NPPs: Overview and general comparison for Paks NPP

    International Nuclear Information System (INIS)

    Guerpinar, A.; Zola, M.

    2001-01-01

    Within the framework of the IAEA coordinated 'Benchmark Study for the seismic analysis and testing of WWER-type NPPs', in-situ dynamic structural testing activities have been performed at the Paks Nuclear Power Plant in Hungary. The specific objective of the investigation was to obtain experimental data on the actual dynamic structural behaviour of the plant's major constructions and equipment under normal operating conditions, for enabling a valid seismic safety review to be made. This paper refers on the comparison of the results obtained from the experimental activities performed by ISMES with those coming from analytical studies performed for the Coordinated Research Programme (CRP) by Siemens (Germany), EQE (Bulgaria), Central Laboratory (Bulgaria), M. David Consulting (Czech Republic), IVO (Finland). This paper gives a synthetic description of the conducted experiments and presents some results, regarding in particular the free-field excitations produced during the earthquake-simulation experiments and an experiment of the dynamic soil-structure interaction global effects at the base of the reactor containment structure. The specific objective of the experimental investigation was to obtain valid data on the dynamic behaviour of the plant's major constructions, under normal operating conditions, to support the analytical assessment of their actual seismic safety. The full-scale dynamic structural testing activities have been performed in December 1994 at the Paks (H) Nuclear Power Plant. The Paks NPP site has been subjected to low level earthquake-like ground shaking, through appropriately devised underground explosions, and the dynamic response of the plant's 1st reactor unit important structures was appropriately measured and digitally recorded, with the whole nuclear power plant under normal operating conditions. In-situ free field response was measured concurrently and, moreover, site-specific geophysical and seismological data were simultaneously

  12. Dbo/Henji Modulates Synaptic dPAK to Gate Glutamate Receptor Abundance and Postsynaptic Response.

    Directory of Open Access Journals (Sweden)

    Manyu Wang

    2016-10-01

    Full Text Available In response to environmental and physiological changes, the synapse manifests plasticity while simultaneously maintains homeostasis. Here, we analyzed mutant synapses of henji, also known as dbo, at the Drosophila neuromuscular junction (NMJ. In henji mutants, NMJ growth is defective with appearance of satellite boutons. Transmission electron microscopy analysis indicates that the synaptic membrane region is expanded. The postsynaptic density (PSD houses glutamate receptors GluRIIA and GluRIIB, which have distinct transmission properties. In henji mutants, GluRIIA abundance is upregulated but that of GluRIIB is not. Electrophysiological results also support a GluR compositional shift towards a higher IIA/IIB ratio at henji NMJs. Strikingly, dPAK, a positive regulator for GluRIIA synaptic localization, accumulates at the henji PSD. Reducing the dpak gene dosage suppresses satellite boutons and GluRIIA accumulation at henji NMJs. In addition, dPAK associated with Henji through the Kelch repeats which is the domain essential for Henji localization and function at postsynapses. We propose that Henji acts at postsynapses to restrict both presynaptic bouton growth and postsynaptic GluRIIA abundance by modulating dPAK.

  13. Practice of fuel management and outage strategy at Paks NPP

    International Nuclear Information System (INIS)

    Farago, P.; Hamvas, I.; Szecsenyi, Zs.; Nemes, I.; Javor, E.

    2000-01-01

    The Paks Nuclear Power Plant generates almost 40% of Hungarian electricity production at lowest price. In spite of this fact the reduction of operational and maintenance costs is one of the most important goal of the plant management. The proper fuel management and outage strategy can give a considerable influence for this cost reduction. The aim of loading pattern planning is to get the required cycle length with available fuel cassettes and to keep all key parameters of safety analysis under safety limits. Another important point is production at profit, where both the fuel and spent fuel cost are determining. Earlier the conditions given by our only fuel supplier restricted our possibilities, so at the beginning the fuel arrangement changing was the only way to improve efficiency of fuel using. As first step we introduced the low leakage core design. The next step was the 4 years cycle using of some cassettes. By this way nearly half of 3 years cycle old cassettes remained in the core for fourth cycle. In the immediate future we want to use profiled cassettes developed by Russian supplier. Simultaneously we will load new type of WWER cassettes with burnable poison developed by BNFL Company. Hereby we can apply more BNFL cassettes for four years cycle even more. Both cost of fuel and number of spent fuel can be reduced besides keeping parameters under safety limits. The Hungarian in service inspection rules determine that every four year we have to make a complete inspection of reactor vessel. Therefore earlier we had two types of outages. Every 4 years we planned a long outage with 55-65 days duration and normal ones with about 30-35 days duration between the long ones. During the normal outages this way did not give us enough room to utilise the shortest possible critical path determined by works on reactor. Some years ago we changed our outage strategy. Now we plan every 4 years a long outage, and between them one normal and two short ones. As a result the

  14. Improving Research Reactor Accident Response Capability at the Hungarian Nuclear Safety Authority

    International Nuclear Information System (INIS)

    Vegh, J.; Gajdos, F.; Horvath, Cs.; Matisz, A.; Nyisztor, D.

    2013-06-01

    The paper describes the design and implementation of an on-line operation monitoring and accident response support system to be used at the CERTA emergency response centre of Hungarian Atomic Energy Authority (HAEA). The monitored facility is the Budapest Research Reactor (BRR), which is a tank-type thermal reactor having 10 MW thermal power. The basic reason for the development of the on-line safety information system is to extend the emergency response capability of the CERTA crisis centre to include emergencies related to BRR, as well. CERTA is operated by HAEA at its Budapest headquarters and the centre already has an on-line system for monitoring the state of the four units of Paks NPP, Hungary. The system is called CERTA VITA and it is able to monitor the four VVER-440/V213 units during their normal operation, and during emergencies (including severe accidents). Ensuring appropriate emergency response capabilities, as well as improving the presently available systems and tools was one of the important recommendations resulting from the analyses following the severe accident at Fukushima. This task is valid not only for the operators of the nuclear facilities but also for the nuclear safety authorities, therefore HAEA decided to launch a project - together with the Centre for Energy Research, the operator of BRR - to establish an on-line information system similar to the CERTA VITA used for monitoring the four units of the Paks NPP. It is believed that by the introduction of this new on-line system the accident response capabilities of HAEA will be further enhanced and the BRR emergencies will be handled at the same professional level as potential emergencies at Paks NPP. (authors)

  15. Example of an application of systematic approach to training. Paks NPP Ltd

    International Nuclear Information System (INIS)

    Szabo, Z.

    1998-01-01

    In order to show a practical example of how Hungarian Maintenance program works implementation of SAT based training of primary circuit valve maintenance senior mechanic is described and explained in detail as done at Paks NPP

  16. Method for investigation of various iodine species in the primary coolant of the nuclear power plant in Paks

    International Nuclear Information System (INIS)

    Volent, G.; Gimesi, O.; Solymosi, J.

    1996-01-01

    Iodine isotopes formed in the course of fission in nuclear reactors may be present in the primary coolant in different oxidation states, i.e., in different chemical forms. It is important to know the chemical forms and their proportions in order to asses the environmental effect of the emitted iodine and the performance of air filters used in the primary circuit for binding iodine, species, since both depend on the chemical forms in which it is present. Volatile components were separated from water samples taken separately from each block of the nuclear power station by purging with inert gas, then the aerosol, iodine vapour and alkyl iodides were selectively bound on the filter system of the 'KOMBI' sampler. I 3 - , I - , IO - , IO 3 - and IO 4 - left in the aqueous phase after purging were separated by consecutive physical and chemical procedures (extraction, isotope exchange, reduction). The results of the investigations have shown that the water technology used in the Nuclear Power Plant in Paks is appropriate with respect to the radioiodine balance. Iodine was found to be predominant species, and no volatile iodine species were found to be present in the primary coolant. Volatile iodine species sometimes appearing in emissions may be formed from leaching waters due to secondary effects. (author)

  17. Blind pre-analysis of the main building complex WWER-440/213 Paks for comparison of analytical and experimental results obtained by explosive testing (task 7a of workplan 95/96)

    International Nuclear Information System (INIS)

    1999-01-01

    Within the research programme on Benchmark studies of seismic analysis of WWER type reactors the blind pre-analysis must be prepared for the main building complex of Paks NPP, based on given excitations derived from explosion tests. The aim of the investigation was to validate different idealization concepts (mathematical models for the idealization of the structures and the soil) as well as investigation procedures (time domain and frequency domain analysis) and finally the software tools by comparing dynamic properties (eigenfrequencies, eigenmodes, modal values) and structural response results (time histories and response spectra). This report contains results of the blind pre-analysis performed by using three dimensional idealization of the main building complex (reactor building, turbine house, galleries) by means of time and frequency domian calculation procedures

  18. Comparing the influence of selenite (Se4+) and selenate (Se6+) on the inhibition of the mercury (Hg) phytotoxicity to pak choi.

    Science.gov (United States)

    Tran, Thi Anh Thu; Dinh, Quang Toan; Cui, Zeiwei; Huang, Jie; Wang, Dan; Wei, Tianjiao; Liang, Dongli; Sun, Xin; Ning, Ping

    2018-01-01

    Selenite (Se (IV)) and selenate (Se (IV)) have recently been demonstrated to be equally effective in inhibiting mercury (Hg) phytotoxicity to plants. This assertion is still unclear. In this study, we aimed to explore the potential effects of Se species (Se 4+ and Se 6+ ) on the inhibition of the mercury (Hg) bioavailability to pak choi in dry land. Pot experiments with exposure to different dosages of mercuric chloride (HgCl 2 ) and selenite (Na 2 SeO 3 ) or selenate (Na 2 SeO 4 ) were treated. To compare the influence of Se (IV) and Se (VI) on the bioaccumulation and bioavailability of Hg, the levels of total Hg in different pak choi (Brassica chinensis L.) tissues (roots and shoots) and the distribution changes of Hg fractions in soil before planting and after harvest were determined as well as the Hg I R values in soils (relative binding intensity) were analyzed. Results showed that application Se (IV) reduced the concentrations of Hg in pak choi roots more than Se (VI). Hg concentrations were also decreased in pak choi shoots in Se (IV) treatments, while which notably increased in Se (VI) treatments. Thus, Se (IV) plays a more important role than Se (VI) in limiting the absorption and bioaccumulation of Hg in pak choi. Moreover, this inhibition may only significantly occur when Se (IV) is at an appropriate level (2.5mg/kg). In addition, the good correlations between the proportions of mobile Hg fractions (soluble and exchangeable fractions), I R values with the Hg concentrations in plants were observed. This affirmed the importance of the Hg fractions transformation and the I R indicator of Hg in the assessment of their bioavailability. Our findings regarding the importance of Se (IV) influence in reducing Hg bioaccumulation not only provided the correct appraisal about the effect of Se species on the inhibition of the Hg phytotoxicity to pak choi in dry land, but also be a good reference for selecting Se fertilizer forms (Se 4+ or Se 6+ ). Copyright © 2017

  19. Pak2 Controls Acquisition of NKT Cell Fate by Regulating Expression of the Transcription Factors PLZF and Egr2.

    Science.gov (United States)

    O'Hagan, Kyle L; Zhao, Jie; Pryshchep, Olga; Wang, Chyung-Ru; Phee, Hyewon

    2015-12-01

    NKT cells constitute a small population of T cells developed in the thymus that produce large amounts of cytokines and chemokines in response to lipid Ags. Signaling through the Vα14-Jα18 TCR instructs commitment to the NKT cell lineage, but the precise signaling mechanisms that instruct their lineage choice are unclear. In this article, we report that the cytoskeletal remodeling protein, p21-activated kinase 2 (Pak2), was essential for NKT cell development. Loss of Pak2 in T cells reduced stage III NKT cells in the thymus and periphery. Among different NKT cell subsets, Pak2 was necessary for the generation and function of NKT1 and NKT2 cells, but not NKT17 cells. Mechanistically, expression of Egr2 and promyelocytic leukemia zinc finger (PLZF), two key transcription factors for acquiring the NKT cell fate, were markedly diminished in the absence of Pak2. Diminished expression of Egr2 and PLZF were not caused by aberrant TCR signaling, as determined using a Nur77-GFP reporter, but were likely due to impaired induction and maintenance of signaling lymphocyte activation molecule 6 expression, a TCR costimulatory receptor required for NKT cell development. These data suggest that Pak2 controls thymic NKT cell development by providing a signal that links Egr2 to induce PLZF, in part by regulating signaling lymphocyte activation molecule 6 expression. Copyright © 2015 by The American Association of Immunologists, Inc.

  20. Results and interpretation of noise measurements using in-core self powered neutron detector strings at Unit 2 of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gloeckler, O.; Por, G.; Valko, J.

    1986-11-01

    In-core neutron noise and fuel assembly outlet temperature noise measurements were performed at Unit 2 of Paks Nuclear Power Plant. Characteristics of the reactor and the noise measuring equipment are briefly described. The in-core Rhodium emitter selfpowered neutron detector strings positioned axially above the other show high coherence and linear phase at low frequencies indicating a marked transport effect, not regularly measured in PWRs. The coherence between horizontally placed neutron detectors is small and the phase is zero. A transport effect of different nature is obtained between neutron detectors (in-core and ex-core) and fuel assembly outlet thermocouples. The observed characteristics depend on reactor and fuel assembly power in a way supporting interpretation in terms of coolant density and void content changes and power feedback effects. During routine analysis vibration of 1.1 Hz appeared as a strong peak in the power spectra. The control assembly that was responsible for the observed behaviour could be localized with high certainty. (author)

  1. OECD-IAEA Paks Fuel Project. Detailed Description of the Results of Calculations

    International Nuclear Information System (INIS)

    2010-05-01

    On 10 April 2003 severe damage of fuel assemblies took place during an incident at Unit 2 of Paks Nuclear Power Plant in Hungary. The assemblies were being cleaned in a special tank below the water level of the spent fuel storage pool in order to remove crud buildup. That afternoon, the chemical cleaning of assemblies was completed and the fuel rods were being cooled by circulation of storage pool water. The first sign of fuel failure was the detection of some fission gases released from the cleaning tank during that evening. The cleaning tank cover locks were released after midnight and this operation was followed by a sudden increase in activity concentrations. The visual inspection revealed that all 30 fuel assemblies were severely damaged. The first evaluation of the event showed that the severe fuel damage happened due to inadequate coolant circulation within the cleaning tank. The damaged fuel assemblies will be removed from the cleaning tank in 2005 and will be stored in special canisters in the spent fuel storage pool of the Paks NPP. Following several discussions between expert from different countries and international organisations the OECD-IAEA Paks Fuel Project was proposed. The project is envisaged in two phases. - Phase 1 is to cover organization of visual inspection of material, preparation of database, performance of analyses and preparatory work for fuel examination. - Phase 2 is to cover the fuel transport and the hot cell examination

  2. Description of Website for the OECD-IAEA Paks Fuel Project

    International Nuclear Information System (INIS)

    Szabo, Emese; Hozer, Zoltan; Nagy, Imre

    2010-01-01

    The first version of the database for the OECD-IAEA PAKS FUEL PROJECT has been collected and it is available on the following password protected website: http://nagy.aeki.kfki.hu Several modifications have been made and new items added according to the minutes of the 1st meeting held in Budapest on 30-31 January 2005

  3. Dynamics of TRIGA-3 Salazar Reactor

    International Nuclear Information System (INIS)

    Gallardo S, L.F.

    1990-01-01

    The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

  4. Leveraging the Pre-DFG Residue Thr-406 To Obtain High Kinase Selectivity in an Aminopyrazole-Type PAK1 Inhibitor Series.

    Science.gov (United States)

    Rudolph, Joachim; Aliagas, Ignacio; Crawford, James J; Mathieu, Simon; Lee, Wendy; Chao, Qi; Dong, Ping; Rouge, Lionel; Wang, Weiru; Heise, Christopher; Murray, Lesley J; La, Hank; Liu, Yanzhou; Manning, Gerard; Diederich, François; Hoeflich, Klaus P

    2015-06-11

    To increase kinase selectivity in an aminopyrazole-based PAK1 inhibitor series, analogues were designed to interact with the PAK1 deep-front pocket pre-DFG residue Thr-406, a residue that is hydrophobic in most kinases. This goal was achieved by installing lactam head groups to the aminopyrazole hinge binding moiety. The corresponding analogues represent the most kinase selective ATP-competitive Group I PAK inhibitors described to date. Hydrogen bonding with the Thr-406 side chain was demonstrated by X-ray crystallography, and inhibitory activities, particularly against kinases with hydrophobic pre-DFG residues, were mitigated. Leveraging hydrogen bonding side chain interactions with polar pre-DFG residues is unprecedented, and similar strategies should be applicable to other appropriate kinases.

  5. Corporate portal system at PAKS NPP, Hungary

    International Nuclear Information System (INIS)

    2009-01-01

    The new Corporate Portal System (CPS) of Paks NPP was launched in November 2006. The portal is based on one of the latest technologies, Plumtree Enterprise WEB 5.0. The main purpose of the installation of the new technology was to serve the working culture change, to give a platform to access all information and applications including the integrated process model used at the NPP. The new technology also supports those goals which were defined in the organization development programme: e.g. to improve internal communication with the establishment of communities of practice. Installation of the CPS has provided a powerful tool for knowledge management; it is possible to share and find all information through a controlled access in documents from various sources, to have links to people, portlets and different communities. Document management of the Paks NPP is supported by the integration of the Document 5 application, as the new Electronic Data Management System (EDMS) and the CPS. Depending on their access rights, all users of the CPS, through Microsoft Internet Explorer, can access technical, economic and human resources documents which are stored anywhere on the internal network (file servers, EDMS, old INRANET). The CPS is also accessible from the internet through a secure connection. The main concept is the integration of all applications to one platform and to help users to find all information they need. An access control list specifies which users and groups have access to an object (and what kind of access privileges they have such as read, select, edit, admin)

  6. Evaluation of the CervidTB STAT-PAK for the detection of Mycobacterium bovis infection in wild deer in Great Britain.

    Science.gov (United States)

    Gowtage-Sequeira, S; Paterson, A; Lyashchenko, K P; Lesellier, S; Chambers, M A

    2009-10-01

    Deer are acknowledged as hosts of Mycobacterium bovis, the causative agent of bovine tuberculosis (bTB), and determining the prevalence of infection in deer species is one of the key steps in understanding the epidemiological role played by cervids in the transmission and maintenance of bTB in the United Kingdom. This study evaluated a rapid lateral-flow test for the detection of bTB in samples from wild deer species in the United Kingdom. Fallow deer (Dama dama), roe deer (Capreolus capreolus), and red deer (Cervus elaphus) from areas in Wales, the Cotswolds, and southwestern England were necropsied for a bTB survey. Serum samples from individual deer were tested with the CervidTB STAT-PAK, and the results were evaluated against the culture of M. bovis from tissues (n = 432). Sensitivity and specificity were 85.7% (95% confidence interval [CI], 42.1 to 99.6%) and 94.8% (95% CI, 92.3 to 96.7%), respectively, with an odds ratio of 109.9 (95% CI, 12.7 to 953.6%) for a positive STAT-PAK result among culture-positive deer. The low prevalence of infection (3.8%, n = 860) affected the confidence of the sensitivity estimate of the test, but all culture-positive fallow deer (n = 6) were detected by the test. In addition, antibodies to M. bovis could be detected in poor-quality serum samples. The results suggest that the CervidTB STAT-PAK could be deployed as a field test for further evaluation.

  7. Rac1-dependent recruitment of PAK2 to G 2 phase centrosomes and their roles in the regulation of mitotic entry

    DEFF Research Database (Denmark)

    May, Martin; Schelle, Ilona; Brakebusch, Cord Herbert

    2014-01-01

    -GTPases Rac/Cdc42. In this study, Rac1 (but not RhoA or Cdc42) is presented to associate with the centrosomes from early G 2 phase until prometaphase in a cell cycle-dependent fashion, as evidenced by western blot analysis of prepared centrosomes and by immunolabeling. PAK associates with the G 2/M......-phase centrosomes in a Rac1-dependent fashion. Furthermore, specific inhibition of Rac1 by C. difficile toxinB-catalyzed glucosylation or by knockout results in inhibited activation of PAK1/2, Aurora A, and the CyclinB/Cdk1 complex in late G 2 phase/prophase and delayed mitotic entry. Inhibition of PAK activation...

  8. Glutamine nitrogen and ammonium nitrogen supplied as a nitrogen source is not converted into nitrate nitrogen of plant tissues of hydroponically grown pak-choi (Brassica chinensis L.).

    Science.gov (United States)

    Wang, H-J; Wu, L-H; Tao, Q-N; Miller, D D; Welch, R M

    2009-03-01

    Many vegetables, especially leafy vegetables, accumulate NO(-) (3)-N in their edible portions. High nitrate levels in vegetables constitute a health hazard, such as cancers and blue baby syndrome. The aim of this study was to determine if (1) ammonium nitrogen (NH(+) (4)-N) and glutamine-nitrogen (Gln-N) absorbed by plant roots is converted into nitrate-nitrogen of pak-choi (Brassica chinensis L.) tissues, and (2) if nitrate-nitrogen (NO(-) (3)-N) accumulation and concentration of pak-choi tissues linearly increase with increasing NO(-) (3)-N supply when grown in nutrient solution. In experiment 1, 4 different nitrogen treatments (no nitrogen, NH(+) (4)-N, Gln-N, and NO(-) (3)-N) with equal total N concentrations in treatments with added N were applied under sterile nutrient medium culture conditions. In experiment 2, 5 concentrations of N (from 0 to 48 mM), supplied as NO(-) (3)-N in the nutrient solution, were tested. The results showed that Gln-N and NH(+) (4)-N added to the nutrient media were not converted into nitrate-nitrogen of plant tissues. Also, NO(-) (3)-N accumulation in the pak-choi tissues was the highest when plants were supplied 24 mM NO(-) (3)-N in the media. The NO(-) (3)-N concentration in plant tissues was quadratically correlated to the NO(-) (3)-N concentration supplied in the nutrient solution.

  9. A conceptual magnetic fabric development model for the Paks loess in Hungary

    DEFF Research Database (Denmark)

    Bradák, B.; Ujvari, Gabor; Seto, Y.

    2018-01-01

    We describe magnetic fabric and depositional environments of aeolian (loess) deposits from Paks, Hungary, and develop a novel, complex conceptual sedimentation model based on grain size and low-field magnetic susceptibility anisotropy data. A plot of shape factor (magnetic fabric parameter) and d...

  10. 123I-Iodomethyl tyrosine radiochemical synthesis and quantification of residual impurities after SepPak purification

    International Nuclear Information System (INIS)

    Matte, G.; Abrams, D.; Kumar, P.; Mercer, J.

    2002-01-01

    [123-I]-Iodomethyl tyrosine, an analog of tyrosine, is used as a radiopharmaceutical to detect malignant tissue in vivo. Initial synthesis report removal of the starting material using HPLC reversed phase chromatography as well as a simple method using a C18-SepPak cartridge when an HPLC system is not available. Small amounts of residual starting material have not been reported to interfere with tumor uptake following biodistribution in vivo. However, in vitro tissue culture studies do require the final product to be free of un-reacted methyl tyrosine. Our goal was to quantify the amount of residual methyl tyrosine after C18-SepPak purification to confirm that 123 I methyl tyrosine purified in this manner would be suitable tissue culture studies. A preconditioned (rinsed with 2 ml Ethanol and 8 ml PBS) C18-SepPak cartridge is loaded with the 123 I-iodomethyl tyrosine reaction mixture and washed with 8 mL of PBS to remove the un-reacted methyl tyrosine and free 123 I-iodide. The cartridge is then eluted with a 20% alcohol/PBS mixture to recover the 123 I-iodomethyl tyrosine. Paper chromatography confirmed the removal of un-reacted 123 I-iodide. A parallel study with a methyl tyrosine standard was used to confirm the removal of the methyl tyrosine from the SepPak cartridge during the washing with 8 mL of PBS. Fractions were collected and UV absorbance was recorded. A standard curve was prepared using the UV absorbance of serial dilutions of methyl tyrosine. The detection limit was in the order of ng/mL. An elution profile of both 123 I methyl tyrosine and methyl tyrosine was obtained and shows that traces of methyl tyrosine can still be present after an 8 mL PBS wash

  11. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 1. Data related to sites and plants: Paks NPP, Kozloduy NPP. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to sites and NPPs Paks and Kozloduy

  12. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 1. Data related to sites and plants: Paks NPP, Kozloduy NPP. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to sites and NPPs Paks and Kozloduy.

  13. Evaluation of the CervidTB STAT-PAK for the Detection of Mycobacterium bovis Infection in Wild Deer in Great Britain▿

    Science.gov (United States)

    Gowtage-Sequeira, S.; Paterson, A.; Lyashchenko, K. P.; Lesellier, S.; Chambers, M. A.

    2009-01-01

    Deer are acknowledged as hosts of Mycobacterium bovis, the causative agent of bovine tuberculosis (bTB), and determining the prevalence of infection in deer species is one of the key steps in understanding the epidemiological role played by cervids in the transmission and maintenance of bTB in the United Kingdom. This study evaluated a rapid lateral-flow test for the detection of bTB in samples from wild deer species in the United Kingdom. Fallow deer (Dama dama), roe deer (Capreolus capreolus), and red deer (Cervus elaphus) from areas in Wales, the Cotswolds, and southwestern England were necropsied for a bTB survey. Serum samples from individual deer were tested with the CervidTB STAT-PAK, and the results were evaluated against the culture of M. bovis from tissues (n = 432). Sensitivity and specificity were 85.7% (95% confidence interval [CI], 42.1 to 99.6%) and 94.8% (95% CI, 92.3 to 96.7%), respectively, with an odds ratio of 109.9 (95% CI, 12.7 to 953.6%) for a positive STAT-PAK result among culture-positive deer. The low prevalence of infection (3.8%, n = 860) affected the confidence of the sensitivity estimate of the test, but all culture-positive fallow deer (n = 6) were detected by the test. In addition, antibodies to M. bovis could be detected in poor-quality serum samples. The results suggest that the CervidTB STAT-PAK could be deployed as a field test for further evaluation. PMID:19656989

  14. Preapplication safety evaluation report for the Sodium Advanced Fast Reactor (SAFR) liquid-metal reactor

    International Nuclear Information System (INIS)

    King, T.L.; Landry, R.R.; Throm, E.D.; Wilson, J.N.

    1991-12-01

    This safety evaluation report (SER) presents the final results of a preapplication design review for the Sodium Advanced Fast Reactor (SAFR) liquid metal reactor (Project 673). The SAFR conceptual design was submitted by the US Department of Energy (DOE) in accordance with the US Nuclear Regulatory Commission (NRC) ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 FR 24643 which provides for the early Commission review and interaction). The standard SAFR plant design consists of four identical reactor modules, referred to as ''paks,'' each with a thermal output rating of 900 MWt, coupled with four steam turbine-generator sets. The total electrical output was held to be 1400 MWe. This SER represents the NRC staff's preliminary technical evaluation of the safety features in the SAFR design. It must be recognized that final conclusions in all matters discussed in this SER require approval by the Commission. During the NRC staff review of the SAFR conceptual design, DOE terminated work on this design in September 1988. This SER documents the work done to that date and no additional work is planned for the SAFR

  15. Iron deficiency anemia interfering the diagnosis of compound heterozygosity for Hb constant spring and Hb Paksé: The first case report.

    Science.gov (United States)

    Chiasakul, Thita; Uaprasert, Noppacharn

    2018-01-01

    Diagnosis of thalassemia or hemoglobinopathy concomitant with iron deficiency anemia (IDA) is challenging. We report a case of 43-year-old female whose diagnosis of compound heterozygosity for hemoglobin Constant Spring (HbCS) and Hb Paksé became apparent after the treatment of IDA. Prior to treatment, Hb analysis using isoelectric focusing (IEF) showed HbA 95.6%, HbA 2 2.7%, and HbCS 1.7% compatible with heterozygous HbCS. After 4 months of oral iron therapy resulting in an improved Hb level, her HbCS level was substantially increased to 8.7% on IEF suggesting homozygous HbCS. Subsequent DNA analysis using multiplex amplification refractory mutation system analysis revealed compound heterozygosity for HbCS and Hb Paksé. This case demonstrated that IDA can significantly reduce HbCS/Hb Paksé levels and probably mask the diagnosis of homozygous HbCS, homozygous Hb Paksé or the compound heterozygosity for both hemoglobinopathies by hemoblogin analysis. The test should be repeated after resolution of IDA, or molecular testing should be performed to confirm the diagnosis. © 2017 Wiley Periodicals, Inc.

  16. Vertical Population Gradients in NGC 891. I. ∇Pak Instrumentation and Spectral Data

    Science.gov (United States)

    Eigenbrot, Arthur; Bershady, Matthew A.

    2018-02-01

    We have measured vertical and radial stellar population gradients in NGC 891. We compare these gradients to those known for the Milky Way from studies of resolved stars. Optical spectroscopic measurements extend spatially from the disk midplane up to 2.6 {kpc} in height and out to a radius of 12 {kpc} on both sides of the galaxy. Data were acquired with ∇Pak, a variable-pitch fiber integral field unit (IFU) on the WIYN telescope. We describe the laboratory and on-sky performance of ∇Pak, as well as modifications to the standard observational and analysis procedures necessary to calibrate data taken with this unique IFU. ∇Pak has a mean throughput of 80% at 5500 \\mathringA . To achieve an estimated precision of 10% in light-weighted mean age and metallicity, we define a set of spatial apertures in radius and height in which spectra are binned to achieve a signal-to-noise ratio of ∼20 Å‑1. We use spectral indices to measure age, metallicity, and abundance, indicating that NGC 891's stellar populations have 0.2 7 {Gyr}) stellar populations at 0.4 {kpc}, roughly the scale height of the thin disk. We also find a slight trend toward younger populations at larger radii, consistent with flaring in an inside-out disk formation scenario. The vertical age gradient in NGC 891 is in remarkable qualitative agreement with a model for disk heating tuned to studies of the Milk Way’s solar cylinder.

  17. Simulation of the Paks-2 incident. The CODEX-CT-1 experiment

    International Nuclear Information System (INIS)

    Windberg, P.; Hozer, Z.; Nagy, I.; Vimi, A.

    2006-01-01

    The Paks-2 cleaning tank incident was simulated with an electrically heated fuel bundle in the CODEX facility. The test conditions included seven hours of oxidation in hydrogen rich steam and final water quenching of the brittle fuel rods. The final state of the bundle showed similar picture that was observed after the incident at the power plant in 2003. (author)

  18. Analyses for licensing of new fuel types at Paks NPP

    International Nuclear Information System (INIS)

    Kereszturi, A.; Bogatyr, S.; Miko, S.; Nemes, I.

    2003-01-01

    In the last years Paks NPP initiated several projects aiming at the introduction of new fuel types and resulting in more economic fuel cycles. The motivations, the reasons, and the economic consequences of the above modifications are detailed. The application of a new fuel type requires the renewal of the relevant chapters of the Safety Analysis Report. The fulfilment of fuel design basis requirements, to be summarised briefly also in the paper, must be investigated during normal and accidental conditions. The characteristics of the different codes, the data transfer between them are detailed. After, the cases of the Normal Operation, Anticipated Operation Occurrence, and the Postulated Accidents, judged as the most relevant ones in case of fuel modifications, are overviewed. In the last part, selected examples of the licensing calculations, performed by the above tools are presented. In conclusion, modifications of the WWER fuel, namely increased enrichment, application of burnable fuel pins, modified geometry make more economic fuel cycles (larger discharge burnup, power up-rate, reduced pressure vessel fluence) are possible. The further step (increased enrichment, burnable poison) of the fuel modernisation at NPP Paks is necessary for more economic fuel cycles and fuel consuming. A sound basis of licensing methodology, safety analysis, and necessary computer codes for the WWER fuel modernisation is available

  19. Final evaluation report for Lockheed Idaho Technologies Company, ARROW-PAK packaging, Docket 95-40-7A, Type A container

    International Nuclear Information System (INIS)

    Kelly, D.L.

    1995-11-01

    The report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the ARROW-PAK packaging. The ARROW-PAK packaging system consists of Marlex M-8000 Driscopipe (Series 8000 [gas] or Series 8600 [industrial]) resin pipe, manufactured by Phillips-Driscopipe, Inc., and is sealed with two dome-shaped end caps manufactured from the same materials. The patented sealing process involves the use of electrical energy to heat opposing faces of the pipe and end caps, and hydraulic rams to press the heated surfaces together. This fusion process produces a homogeneous bonding of the end cap to the pipe. The packaging may be used with or without the two internal plywood spacers. This packaging was evaluated and tested in October 1995. The packaging configuration described in this report is designed to ship Type A quantities of solid radioactive materials, Form No. 1, Form No. 2, and Form No. 3

  20. Test Plan for Lockheed Idaho Technologies Company (LITCO), ARROW-PAK Packaging, Docket 95-40-7A, Type A Container

    International Nuclear Information System (INIS)

    Kelly, D.L.

    1995-01-01

    This report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance testing to be followed for qualification of the Lockheed Idaho Technologies Company, ARROW-PAK, for use as a Type A Packaging. The packaging configuration being tested is intended for transportation of radioactive solids, Form No. 1, Form No. 2, and Form No. 3

  1. Impact of Heat-Shock Treatment on Yellowing of Pak Choy Leaves

    Institute of Scientific and Technical Information of China (English)

    WANG Xiang-yang; SHEN Lian-qing; YUAN Hai-na

    2004-01-01

    The physiological mechanism of maintaining the green colour of pak choy leaves (Brassica rapa var chinensis) with heat-shock treatment was studied. Chlorophyll in the outer leaves of pak choy degraded rapidly during storage at ambient temperature (20 ± 2℃), a slight yellow appeared. Heat-shock treatment (46- 50℃) had a mild effect on maintaining the green colour of outer leaves. Normal chlorophyll degradation was associated with a binding of chlorophyll with chlorophyll-binding-protein preceding chlorophyll breakdown.Heat-shock treatment was found to reduce the binding-capacity between chlorophyllbinding-protein and chlorophyll. In the chlorophyll degradation pathway, pheide dioxygenase was synthesized during leaf senescence which was considered to be a key enzyme in chlorophyll degradation. Activity of this enzyme was reduced following heat-shock treatment, which might explain the observed reduction in chlorophyll breakdown. Two groups of heat-shock proteins were detected in treated leaves, the first group containing proteins from 54KDa to 74 Kda, and the second group contained proteins from 15 KDa to 29KDa. Heat-shock treatment was also found to retard the decline of glucose and fructose (the main energy substrates) of outer leaves.

  2. Development and installation of a new on-line plant safety monitoring system for the Paks VVER-440 units

    International Nuclear Information System (INIS)

    Vegh, J.; Major, C.; Buerger, L.; Lipcsei, S.; Horvath, C.; Kapocs, G.; Eiler, J.; Hornaes, A.; Hulsund, J.E.

    2000-01-01

    The paper describes the architecture, modules, algorithms and human-machine interface of a new operator support system (OSS), which is integrated into the new, reconstructed Paks NPP plant computers. The main task of the new OSS is to perform continuous plant safety monitoring and assessment, it has the following basic functions: on-line evaluation and presentation of critical safety function (CSF) status trees, continuous evaluation and presentation of the actual safety status of the plant, displaying and browsing the new symptom-oriented EOPs, automatic displaying of those process signals which are quoted in the EOPs. The first version of the new operator support system was connected to the Paks NPP full scope simulator in October 1999. This configuration was later successfully applied for the simulator testing of the new symptom-oriented EOP set for the Paks NPP in November 1999. The installation process was continued in 2000: the new system started its operation on Unit 2 (June) and on Unit 1 (August), together with the reconstructed, new PCS. (author)

  3. Pollution and pollution tolerance in the case of polycyclic aromatic hydrocarbons (PAH); Belastung durch Polyzyklische aromatische Kohlenwasserstoffe (PAK)

    Energy Technology Data Exchange (ETDEWEB)

    Renger, M.; Mekiffer, B. [Technische Univ. Berlin (Germany). Inst. fuer Oekologie-Bodenkunde

    1997-12-31

    The purpose of the present follow-up project was to examine the contamination with polycyclic aromatic hydrocarbons (PAH) of different anthropogenic urban soils including clay soils containing demolition waste, household waste, ash, and residues from a coking plant. A further task was to analyse, or infer from other part-projects, standard soil parameters such as organic carbon content, pH, and anion levels in order to clarify any relationships between PAH contamination and the more easily determinable soil characteristics. Furthermore, the sorption behaviour for PAH of selected anthropogenic urban soils was to be characterised by means of batch experiments. [Deutsch] Im Rahmen des Anschlussvorhabens sollte die Kontamination von anthropogenen Stadtboeden- darunter Truemmerschutt-, Hausmuell-, Asche- sowie Kokereilehmboden- durch polyzyklische aromatische Kohlenwasserstoffe (PAK) untersucht werden. Zusaetzlich sollten die bodenkundlichen Standardparameter Corg, pH-Wert, Anionengehalte und KAKpot analysiert bzw. von den anderen Teilvorhaben uebernommen werden, um Zusammenhaenge zwischen der PAK-Kontamination und relativ leicht zu bestimmenden bodenkundlichen Kennwerten klaeren zu koennen. Das Sorptionsverhalten ausgewaehlter anthropogener Stadtboeden fuer PAK sollte durch Batchversuche charakterisiert werden. (orig./SR)

  4. Identification of flavonoids and hydroxycinnamic acids in pak choi varieties (Brassica campestris L. ssp. chinensis var. communis) by HPLC-ESI-MSn and NMR and their quantification by HPLC-DAD.

    Science.gov (United States)

    Harbaum, Britta; Hubbermann, Eva Maria; Wolff, Christian; Herges, Rainer; Zhu, Zhujun; Schwarz, Karin

    2007-10-03

    Twenty-eight polyphenols (11 flavonoid derivatives and 17 hydroxycinnamic acid derivatives) were detected in different cultivars of the Chinese cabbage pak choi ( Brassica campestris L. ssp. chinensis var. communis) by HPLC-DAD-ESI-MS(n). Kaempferol was found to be the major flavonoid in pak choi, glycosylated and acylated with different compounds. Smaller amounts of isorhamnetin were also detected. A structural determination was carried out by (1)H and (13)C NMR spectroscopy for the main compound, kaempferol-3-O-hydroxyferuloylsophoroside-7-O-glucoside, for the first time. Hydroxycinnamic acid derivatives were identified as different esters of quinic acid, glycosides, and malic acid. The latter ones are described for the first time in cabbages. The content of polyphenols was determined in 11 cultivars of pak choi, with higher concentrations present in the leaf blade than in the leaf stem. Hydroxycinnamic acid esters, particularly malic acid derivatives, are present in both the leaf blade and leaf stem, whereas flavonoid levels were determined only in the leaf blade.

  5. Gas formation in drum waste packages of Paks NPP

    International Nuclear Information System (INIS)

    Molnar, M.; Palcsu, L.; Svingor, E.; Szanto, Z.; Futo, I.; Ormai, P.

    2000-01-01

    Gas composition measurements have been carried out by mass spectrometry analysis of samples taken from the headspace of ten drum waste packages generated and temporarily stored at Paks NPP. Four drums contained compacted solid waste, three drums were filled with grouted (solidified) sludge and three drums contained solid waste without compaction. The drums have been equipped with a special gas outlet system to make repeated sampling possible. Based on the first measurements significant differences in the gas composition and the rate of gas generation among the drums were found. (author)

  6. Unveiling the role of PAK2 in CD44 mediated inhibition of proliferation, differentiation and apoptosis in AML cells

    KAUST Repository

    Aldehaiman, Mansour M.

    2018-04-01

    Acute myeloid leukemia (AML) is a heterogeneous disease characterized by the accumulation of immature nonfunctional highly proliferative hematopoietic cells in the blood, due to a blockage in myeloid differentiation at various stages. Since the success of the differentiation agent, All-trans retinoic acid (ATRA), in the treatment of acute promyelocytic leukemia (APL), much effort has gone into trying to find agents that are able to differentiate AML cells and specifically the leukemic stem cell (LSC). CD44 is a cell surface receptor that is over-expressed on AML cells. When bound to anti-CD44 monoclonal antibodies (mAbs), this differentiation block is relieved in AML cells and their proliferation is reduced. The molecular mechanisms that AML cells undergo to achieve this reversal of their apparent phenotype is not fully understood. To this end, we designed a study using quantitative phosphoproteomics approaches aimed at identifying differences in phosphorylation found on proteins involved in signaling pathways post-treatment with CD44-mAbs. The Rho family of GTPases emerged as one of the most transformed pathways following the treatment with CD44-mAbs. The P21 activated kinase 2(PAK2), a target of the Rho family of GTPases, was found to be differentially phosphorylated in AML cells post-treatment with CD44-mAbs. This protein has been found to possess a role similar to that of a switch that determines whether the cell survives or undergoes apoptosis. Beyond confirming these results by various biochemical approaches, our study aimed to determine the effect of knock down of PAK2 on AML cell proliferation and differentiation. In addition, over-expression of PAK2 mutants using plasmid cloning was also explored to fully understand how levels of PAK2 as well as the alteration of specific phospohorylation sites could alter AML cell responses to CD44-mAbs. Results from this study will be important in determining whether PAK2 could be used as a potential therapeutic target

  7. Time versus frequency domain calculation of the main building complex of the VVER 440/213 NPP PAKS

    International Nuclear Information System (INIS)

    Katona, T.; Ratkai, S.; Halbritter, A.; Krutzik, N.J.; Schuetz, W.

    1995-01-01

    Various dynamic analyses were conducted for the main building complex of the VVER 440/213 PAKS in order to determine the dynamic response and assess the aseismic capacity of this nuclear power plant. Different types of mathematical models for idealizing the soil and the building structures were used. The main goal of the study presented here was to demonstrate the effects of different procedures for consideration of soil-structure interaction on the dynamic response of the structures mentioned above. The analyses were based on appropriate mathematical models of the coupled vibration structures (reactor building, turbine hall, intermediate building structures) and the layered soil. On the basis of this study, it can be concluded that substructure models using frequency-independent impedances and cut-off of modal damping usually provide conservative results. Complex models which allow the soil-soil and the structure or by frequency-dependent impedances) provide more accurate results. The latter approach results in more efficient designs which are not only safe but also economical. (author). 7 refs., 15 figs

  8. Characterisation of the inventory of radioisotopes induced in the biological shield a WWER-440 reactor

    International Nuclear Information System (INIS)

    Feher, S.; Czifrus, Sz.; Zsolnay, E.M.; Szondi, E.

    2001-01-01

    A significant part of the radwaste originating from the decommissioning of NPPs is made up of the activated concrete and steel components of the biological shield. The paper presents the results of studies aimed at the determination of the amount of radionuclides accumulating in the serpentinous and ordinary concrete shield around the WWER-440 reactors of the Paks NPP. For the calculations, the reactor, vessel and shield were modelled in detail both in terms of geometry and material composition. The spatial and energy distribution of the activating neutron spectrum was determined by certain modules of SCALE 4.3 and the code TORT in two and three dimensions, while the activation was calculated using ORIGEN-S for 22 geometrical regions. The results showed that the activity of the concrete structures at final shutdown after 30 years of operation is approximately 50 TBq, which decreases to 20, 12, 1.1 TBq and 27 GBq after 1 month, 1 year, 10 and 100 years, respectively (Authors)

  9. Decommissioning of a small reactor (BR3 reactor, Belgium)

    International Nuclear Information System (INIS)

    Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

    2002-01-01

    Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

  10. Description of training activities and re-training system for nuclear professionals at the Paks Nuclear Power Plant, Hungary

    International Nuclear Information System (INIS)

    Jambrich, I.; Trampus, P.

    1993-01-01

    The nuclear power units of Paks, Hungary, have always been operated by Hungarian personnel, from the very beginning. The operator staff of unit 1 acquired its knowledge primarily outside of the country, but since 1983 the overall training process has been run entirely in Hungary, in Paks. This report gives details of present system of training programme in Hungary. The system of training for professionals builds up in vertically linked modules and is job oriented. It begins with theoretical training, followed by programmed on-the-job training which must successfully be finished before a release onto in-company or authority licensing exams for individual job performance

  11. “Liting it up”: Popular Culture, Indo-Pak Basketball, and South Asian American Institutions

    Directory of Open Access Journals (Sweden)

    Stanley Ilango Thangaraj

    2010-08-01

    Full Text Available South Asian American participants of a co-ethnic basketball league, known as Indo-Pak Basketball, utilized urban basketball vernacular through the phrase “liting it up” to identify individuals scoring points in great numbers. The person “liting it up” becomes visible and receives recognition. Accordingly, I want to “lite up” the scholarship on South Asian America whereby situating South Asian American religious sites and cultural centers as key arenas for “Americanization” through US popular culture. I situate sport as a key element of popular culture through which South Asian American communities work out, struggle through, and contest notions of self. Informed by an Anthropology of Sport, ethnography of South Asian American communities in Atlanta takes place alongside an examination of the North American Indo-Pak Basketball circuit. Accordingly, my findings indicate that such community formation has also taken shape at the intersections of institutions, gender, and sexuality whereby excluding queers, women, and other communities of color.

  12. Seismic assessment and upgrading of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Tamas, K.

    2001-01-01

    A comprehensive programme for seismic assessment and upgrading is currently in progress at Hungary's Paks NPP. The re-evaluation of the site seismic hazard had been already completed. The technology of safe shut down and heat removal is established and the systems and structures relevant for seismic safety are identified. A seismic instrumentation is installed. The pre-earthquake preparedness and post-earthquake actions are elaborated. The methods for seismic capacity assessment are selected. The seismic capacity evaluation and the design of upgrading measures are currently in progress. The easy to perform upgrading covering the most urgent measures had been already performed. (author)

  13. Phosphorylation of Threonine 794 on Tie1 by Rac1/PAK1 Reveals a Novel Angiogenesis Regulatory Pathway.

    Directory of Open Access Journals (Sweden)

    Jessica L Reinardy

    Full Text Available The endothelial receptor tyrosine kinase (RTK Tie1 was discovered over 20 years ago, yet its precise function and mode of action remain enigmatic. To shed light on Tie1's role in endothelial cell biology, we investigated a potential threonine phosphorylation site within the juxtamembrane domain of Tie1. Expression of a non-phosphorylatable mutant of this site (T794A in zebrafish (Danio rerio significantly disrupted vascular development, resulting in fish with stunted and poorly branched intersomitic vessels. Similarly, T794A-expressing human umbilical vein endothelial cells formed significantly shorter tubes with fewer branches in three-dimensional Matrigel cultures. However, mutation of T794 did not alter Tie1 or Tie2 tyrosine phosphorylation or downstream signaling in any detectable way, suggesting that T794 phosphorylation may regulate a Tie1 function independent of its RTK properties. Although T794 is within a consensus Akt phosphorylation site, we were unable to identify a physiological activator of Akt that could induce T794 phosphorylation, suggesting that Akt is not the physiological Tie1-T794 kinase. However, the small GTPase Ras-related C3 botulinum toxin substrate 1 (Rac1, which is required for angiogenesis and capillary morphogenesis, was found to associate with phospho-T794 but not the non-phosphorylatable T794A mutant. Pharmacological activation of Rac1 induced downstream activation of p21-activated kinase (PAK1 and T794 phosphorylation in vitro, and inhibition of PAK1 abrogated T794 phosphorylation. Our results provide the first demonstration of a signaling pathway mediated by Tie1 in endothelial cells, and they suggest that a novel feedback loop involving Rac1/PAK1 mediated phosphorylation of Tie1 on T794 is required for proper angiogenesis.

  14. Main building complex WWER 440/213 upper range design response spectra for soft soil site conditions (Paks)

    International Nuclear Information System (INIS)

    Krutzik, N.

    1996-01-01

    Within the Benchmark studies parallel investigation were prepared for the main building complex of the WWER-440/213 Paks NPP by several participating institutions. The investigations were based on various mathematical models and procedures but all had the same seismological data as input. The calculation methods as well as software tools were different. This report covers the enveloped response results which were the basis for the benchmark studies and which should be used for upgrading of mechanical and electrical components and systems which will follow. These response spectra which consider a certain conservatism namely neglecting the frequency independence of the stiffness and the cut-off of damping values are named 'Upper Range design Benchmark Response Spectra' for the main building of Paks NPP

  15. Main building complex WWER 440/213 upper range design response spectra for soft soil site conditions (Paks)

    Energy Technology Data Exchange (ETDEWEB)

    Krutzik, N [Siemens AG, Power Generation Group (KWU) NDA2, Offenbach (Germany)

    1996-07-01

    Within the Benchmark studies parallel investigation were prepared for the main building complex of the WWER-440/213 Paks NPP by several participating institutions. The investigations were based on various mathematical models and procedures but all had the same seismological data as input. The calculation methods as well as software tools were different. This report covers the enveloped response results which were the basis for the benchmark studies and which should be used for upgrading of mechanical and electrical components and systems which will follow. These response spectra which consider a certain conservatism namely neglecting the frequency independence of the stiffness and the cut-off of damping values are named 'Upper Range design Benchmark Response Spectra' for the main building of Paks NPP.

  16. A population growth trend analysis for Neotricula aperta, the snail intermediate host of Schistosoma mekongi, after construction of the Pak-Mun dam.

    Directory of Open Access Journals (Sweden)

    Stephen W Attwood

    2013-11-01

    Full Text Available The Pak-Mun dam is a controversial hydro-power project on the Mun River in Northeast Thailand. The dam is sited in a habitat of the freshwater snail Neotricula aperta, which is the intermediate host for the parasitic blood-fluke Schistosoma mekongi causing Mekong schistosomiasis in humans in Cambodia and Laos. Few data are available which can be used to assess the effects of water resource development on N. aperta. The aim of this study was to obtain data and to analyze the possible impact of the dam on N. aperta population growth.Estimated population densities were recorded for an N. aperta population in the Mun River 27 km upstream of Pak-Mun, from 1990 to 2011. The Pak-Mul dam began to operate in 1994. Population growth was modeled using a linear mixed model expression of a modified Gompertz stochastic state-space exponential growth model. The N. aperta population was found to be quite stable, with the estimated growth parameter not significantly different from zero. Nevertheless, some marked changes in snail population density were observed which were coincident with changes in dam operation policy.The study found that there has been no marked increase in N. aperta population growth following operation of the Pak-Mun dam. The analysis did indicate a large and statistically significant increase in population density immediately after the dam came into operation; however, this increase was not persistent. The study has provided the first vital baseline data on N. aperta population behavior near to the Pak-Mun dam and suggests that the operation policy of the dam may have an impact on snail population density. Nevertheless, additional studies are required for other N. aperta populations in the Mun River and for an extended time series, to confirm or refine the findings of this work.

  17. HydroPak: concept design and analysis of a packaged cross-flow turbine

    International Nuclear Information System (INIS)

    2004-01-01

    This report summarises the findings of a project to complete the conceptual design and economic optimization of a modular standardised crossflow hydro-turbine. Details are given of the work to date, the comparison of HydroPak cost with conventional micro- and mini-hydro power costs, and the economic advantages of taking the ''packaged'' and ''standardised approaches'' to the design process. The market for mini-hydro turbines is discussed

  18. Cultivar-Specific Changes in Primary and Secondary Metabolites in Pak Choi (Brassica Rapa, Chinensis Group by Methyl Jasmonate

    Directory of Open Access Journals (Sweden)

    Moo Jung Kim

    2017-05-01

    Full Text Available Glucosinolates, their hydrolysis products and primary metabolites were analyzed in five pak choi cultivars to determine the effect of methyl jasmonate (MeJA on metabolite flux from primary metabolites to glucosinolates and their hydrolysis products. Among detected glucosinolates (total 14 glucosinolates; 9 aliphatic, 4 indole and 1 aromatic glucosinolates, indole glucosinolate concentrations (153–229% and their hydrolysis products increased with MeJA treatment. Changes in the total isothiocyanates by MeJA were associated with epithiospecifier protein activity estimated as nitrile formation. Goitrin, a goitrogenic compound, significantly decreased by MeJA treatment in all cultivars. Changes in glucosinolates, especially aliphatic, significantly differed among cultivars. Primary metabolites including amino acids, organic acids and sugars also changed with MeJA treatment in a cultivar-specific manner. A decreased sugar level suggests that they might be a carbon source for secondary metabolite biosynthesis in MeJA-treated pak choi. The result of the present study suggests that MeJA can be an effective agent to elevate indole glucosinolates and their hydrolysis products and to reduce a goitrogenic compound in pak choi. The total glucosinolate concentration was the highest in “Chinese cabbage” in the control group (32.5 µmol/g DW, but indole glucosinolates increased the greatest in “Asian” when treated with MeJA.

  19. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  20. Phytoavailability of Cadmium (Cd) to Pak Choi (Brassica chinensis L.) Grown in Chinese Soils: A Model to Evaluate the Impact of Soil Cd Pollution on Potential Dietary Toxicity

    Science.gov (United States)

    Yang, Xiaoe; Xiao, Wendan; Stoffella, Peter J.; Saghir, Aamir; Azam, Muhammad; Li, Tingqiang

    2014-01-01

    Food chain contamination by soil cadmium (Cd) through vegetable consumption poses a threat to human health. Therefore, an understanding is needed on the relationship between the phytoavailability of Cd in soils and its uptake in edible tissues of vegetables. The purpose of this study was to establish soil Cd thresholds of representative Chinese soils based on dietary toxicity to humans and develop a model to evaluate the phytoavailability of Cd to Pak choi (Brassica chinensis L.) based on soil properties. Mehlich-3 extractable Cd thresholds were more suitable for Stagnic Anthrosols, Calcareous, Ustic Cambosols, Typic Haplustalfs, Udic Ferrisols and Periudic Argosols with values of 0.30, 0.25, 0.18, 0.16, 0.15 and 0.03 mg kg−1, respectively, while total Cd is adequate threshold for Mollisols with a value of 0.86 mg kg−1. A stepwise regression model indicated that Cd phytoavailability to Pak choi was significantly influenced by soil pH, organic matter, total Zinc and Cd concentrations in soil. Therefore, since Cd accumulation in Pak choi varied with soil characteristics, they should be considered while assessing the environmental quality of soils to ensure the hygienically safe food production. PMID:25386790

  1. Phytoavailability of cadmium (Cd) to Pak choi (Brassica chinensis L.) grown in Chinese soils: a model to evaluate the impact of soil Cd pollution on potential dietary toxicity.

    Science.gov (United States)

    Rafiq, Muhammad Tariq; Aziz, Rukhsanda; Yang, Xiaoe; Xiao, Wendan; Stoffella, Peter J; Saghir, Aamir; Azam, Muhammad; Li, Tingqiang

    2014-01-01

    Food chain contamination by soil cadmium (Cd) through vegetable consumption poses a threat to human health. Therefore, an understanding is needed on the relationship between the phytoavailability of Cd in soils and its uptake in edible tissues of vegetables. The purpose of this study was to establish soil Cd thresholds of representative Chinese soils based on dietary toxicity to humans and develop a model to evaluate the phytoavailability of Cd to Pak choi (Brassica chinensis L.) based on soil properties. Mehlich-3 extractable Cd thresholds were more suitable for Stagnic Anthrosols, Calcareous, Ustic Cambosols, Typic Haplustalfs, Udic Ferrisols and Periudic Argosols with values of 0.30, 0.25, 0.18, 0.16, 0.15 and 0.03 mg kg-1, respectively, while total Cd is adequate threshold for Mollisols with a value of 0.86 mg kg-1. A stepwise regression model indicated that Cd phytoavailability to Pak choi was significantly influenced by soil pH, organic matter, total Zinc and Cd concentrations in soil. Therefore, since Cd accumulation in Pak choi varied with soil characteristics, they should be considered while assessing the environmental quality of soils to ensure the hygienically safe food production.

  2. Phytoavailability of cadmium (Cd to Pak choi (Brassica chinensis L. grown in Chinese soils: a model to evaluate the impact of soil Cd pollution on potential dietary toxicity.

    Directory of Open Access Journals (Sweden)

    Muhammad Tariq Rafiq

    Full Text Available Food chain contamination by soil cadmium (Cd through vegetable consumption poses a threat to human health. Therefore, an understanding is needed on the relationship between the phytoavailability of Cd in soils and its uptake in edible tissues of vegetables. The purpose of this study was to establish soil Cd thresholds of representative Chinese soils based on dietary toxicity to humans and develop a model to evaluate the phytoavailability of Cd to Pak choi (Brassica chinensis L. based on soil properties. Mehlich-3 extractable Cd thresholds were more suitable for Stagnic Anthrosols, Calcareous, Ustic Cambosols, Typic Haplustalfs, Udic Ferrisols and Periudic Argosols with values of 0.30, 0.25, 0.18, 0.16, 0.15 and 0.03 mg kg-1, respectively, while total Cd is adequate threshold for Mollisols with a value of 0.86 mg kg-1. A stepwise regression model indicated that Cd phytoavailability to Pak choi was significantly influenced by soil pH, organic matter, total Zinc and Cd concentrations in soil. Therefore, since Cd accumulation in Pak choi varied with soil characteristics, they should be considered while assessing the environmental quality of soils to ensure the hygienically safe food production.

  3. Refraktīvā lēcas ķirurģija pie augstas pakāpes ametropijas

    OpenAIRE

    Šamajeva, Nataļja

    2012-01-01

    Ametropija ir patoloģisks redzes stāvoklis, kad gaismas stari, izejot cauri acs optiskajām vidēm, fokusējas nevis uz tīklenes, bet pirms vai aiz tās, tad apkārtējo pasauli cilvēks redz neasi. Ametropijas veidi ir miopija, hipermetropija, astigmatisms, kā arī presbopija. Izšķir vājas, vidējas un augstas pakāpes ametropiju. Pilnībā sekmīgi izārstēt augstas pakāpes ametropijas var ar dabīgās acu lēcas aizvietošanu ar multifokālo intraokulāro lēcu. Turklāt pacientiem jāņem vērā, ka šai operā...

  4. Molecular evolution, characterization and expression analysis of SnRK2 gene family in Pak-choi (Brassica rapa ssp. chinensis

    Directory of Open Access Journals (Sweden)

    Zhinan eHuang

    2015-10-01

    Full Text Available Abstract: The sucrose non-fermenting 1-related protein kinase 2 (SnRK2 family members are plant-specific serine/threonine kinases that are involved in the plant response to abiotic stress and abscisic acid (ABA-dependent plant development. Further understanding of the evolutionary history and expression characteristics of these genes will help to elucidate the mechanisms of the stress tolerance in Pak-choi, an important green leafy vegetable in China. Thus, we investigated the evolutionary patterns, footprints and conservation of SnRK2 genes in selected plants and later cloned and analyzed SnRK2 genes in Pak-choi. We found that this gene family was preferentially retained in Brassicas after the Brassica-Arabidopsis thaliana split. Next, we cloned and sequenced 13 SnRK2 from both cDNA and DNA libraries of stress-induced Pak-choi, which were under conditions of ABA, salinity, cold, heat, and osmotic treatments. Most of the BcSnRK2s have eight exons and could be divided into three groups. The subcellular localization predictions suggested that the putative BcSnRK2 proteins were enriched in the nucleus. The results of an analysis of the expression patterns of the BcSnRK2 genes showed that BcSnRK2 group III genes were robustly induced by ABA treatments. Most of the BcSnRK2 genes were activated by low temperature, and the BcSnRK2.6 genes responded to both ABA and low temperature. In fact, most of the BcSnRK2 genes showed positive or negative regulation under ABA and low temperature treatments, suggesting that they may be global regulators that function at the intersection of multiple signaling pathways to play important roles in Pak-choi stress responses.

  5. Pak1, adjuvant tamoxifen therapy, and breast cancer recurrence risk in a Danish population-based study

    DEFF Research Database (Denmark)

    Ahern, Thomas P; Cronin-Fenton, Deirdre P; Lash, Timothy L

    2016-01-01

    -/TAM - group. Pak1 cytoplasmic intensity was not associated with breast cancer recurrence in either group (ER+/TAM + ORadj for strong vs. no cytoplasmic staining = 0.91, 95% CI 0.57, 1.5; ER-/TAM - ORadj for strong vs. no cytoplasmic staining = 0.74, 95% CI 0.39, 1.4). Associations between Pak1 nuclear......Background Adjuvant tamoxifen therapy approximately halves the risk of estrogen receptor-positive (ER+) breast cancer recurrence, but many women do not respond to therapy. Observational studies nested in clinical trial populations suggest that overexpression or nuclear localization of p21-activated...... by immunohistochemical staining of primary breast tumors from recurrence cases and matched controls from two breast cancer populations; women diagnosed with ER-positive tumors who received at least one year of tamoxifen therapy (ER+/TAM+), and women diagnosed with ER-negative tumors who survived for at least one year...

  6. Rapid synthesis of maleimide functionalized fluorine-18 labeled prosthetic group using "radio-fluorination on the Sep-Pak" method.

    Science.gov (United States)

    Basuli, Falguni; Zhang, Xiang; Jagoda, Elaine M; Choyke, Peter L; Swenson, Rolf E

    2018-03-25

    Following our recently published fluorine-18 labeling method, "Radio-fluorination on the Sep-Pak", we have successfully synthesized 6-[ 18 F]fluoronicotinaldehyde by passing a solution (1:4 acetonitrile: t-butanol) of its quaternary ammonium salt precursor, 6-(N,N,N-trimethylamino)nicotinaldehyde trifluoromethanesulfonate (2), through a fluorine-18 containing anion exchange cartridge (PS-HCO 3 ). Over 80% radiochemical conversion was observed using 10 mg of precursor within 1 minute. The [ 18 F]fluoronicotinaldehyde ([ 18 F]5) was then conjugated with 1-(6-(aminooxy)hexyl)-1H-pyrrole-2,5-dione to prepare the fluorine-18 labeled maleimide functionalized prosthetic group, 6-[ 18 F]fluoronicotinaldehyde O-(6-(2,5-dioxo-2,5-dihydro-1H-pyrrol-1-yl)hexyl) oxime, 6-[ 18 F]FPyMHO ([ 18 F]6). The current Sep-Pak method not only improves the overall radiochemical yield (50 ± 9%, decay-corrected, n = 9) but also significantly reduces the synthesis time (from 60-90 minutes to 30 minutes) when compared with literature methods for the synthesis of similar prosthetic groups. Published 2018. This article is a U.S. Government work and is in the public domain in the USA.

  7. HydroPak: concept design and analysis of a packaged cross-flow turbine

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This report summarises the findings of a project to complete the conceptual design and economic optimization of a modular standardised crossflow hydro-turbine. Details are given of the work to date, the comparison of HydroPak cost with conventional micro- and mini-hydro power costs, and the economic advantages of taking the ''packaged'' and ''standardised approaches'' to the design process. The market for mini-hydro turbines is discussed.

  8. Implementation of safeguards at modular vault dry store at Paks NPP in Hungary

    International Nuclear Information System (INIS)

    Safar, J.; Czoch, I.; Szoellosi, E.F.; Janov, J.; Sannie, G.; Daniel, G.; Szabo, J.L.

    1999-01-01

    A safeguards system has been implemented at the GEC-Alsthom designed Modular Vault Dry Store at Paks NPP in Hungary without previous safeguards related experience for this type of spent fuel storage. C/S measures and sealing have primary importance. In addition. spent fuel attribute signatures are detected by a fuel transfer monitor at the cask load/unload port. These are complemented with the corresponding accounting measures. (author)

  9. Closed-loop automatic photogrammetry-geodesy information system for the construction of the Paks nuclear power plant

    International Nuclear Information System (INIS)

    Detrekoei, Akos; Eoery, Karacson; Sarkoezy, Ferenc

    1984-01-01

    The stereo photogrammetric data collection, measurement and data processing system operating within the Geodesy Plan of the Paks nuclear power plant is described. The interactive graphic computer system, its functions and operation, together with plotters and displays for the generation of graphic output are presented. (R.P.)

  10. Analysis of a buried pipeline at WWER-440/213 Paks NPP

    International Nuclear Information System (INIS)

    1999-01-01

    According to regulations, all safety structures of a NPP must be designed to withstand loads induced by earthquakes. This applies to safety related underground piping systems. These structures are typically embedded in about 2-3 m of layered soil, and sometimes protected by concrete slabs resting on the ground surface. A rigorous solution for the dynamic response of such a structure would require accounting for nonlinear and three dimensional effects. A non-linear analysis is possible by using specialized computer codes and material models to account for non-linear behaviour of the soil. The aim of this paper was to reanalyze the pipeline of Paks NPP in order to demonstrate its dynamic behaviour interacting with the soil and the connected buildings as well as to determine the dynamic responses and the stresses in typical regions of the pipeline. To perform the numerical analysis a three dimensional finite element code (SSASSI/S) based on 'flexible volume method' was applied

  11. Novel radiosynthesis of PET HSV-tk gene reporter probes [18F]FHPG and [18F]FHBG employing dual Sep-Pak SPE techniques.

    Science.gov (United States)

    Wang, Ji-Quan; Zheng, Qi-Huang; Fei, Xiangshu; Mock, Bruce H; Hutchins, Gary D

    2003-11-17

    Positron emission tomography (PET) herpes simplex virus thymidine kinase (HSV-tk) gene reporter probes 9-[(3-[(18)F]fluoro-1-hydroxy-2-propoxy)methyl]guanine ([(18)F]FHPG) and 9-(4-[(18)F]fluoro-3-hydroxymethylbutyl)guanine ([(18)F]FHBG) were prepared by nucleophilic substitution of the appropriate tosylated precursors with [(18)F]KF/Kryptofix 2.2.2 followed by a quick deprotection reaction and purification with a simplified dual Silica Sep-Pak solid-phase extraction (SPE) method in 15-30% radiochemical yield.

  12. Structural response of Paks NPP WWER-440 MW main building complex to blast input motion. Final report

    International Nuclear Information System (INIS)

    1999-01-01

    The Soviet standard design units WWER-440/213 type installed in Paks NPP were not originally designed for a Safe Shutdown Earthquake. At the time of selection of Paks site on the basis of historical earthquake data was supposed that the maximum earthquake is of grade V according MSK-64 scale. This seismicity level had not required any special measures to account for seismic event effects on the Main Building Complex Structure. Current site seismicity studies reveal that the seismic hazard for the site significantly exceeds the originally estimated. In addition the safety rules and seismic code requirements became more rugged. As a part of the activities to increase the seismic safety of the Paks NPP the study on dynamic behaviour of the Main Building Complex Structure has been performed with support of IAEA. The explosion full scale tests were carried out for determining the dynamic behaviour of the structure and for assessment of the Soil Structure Interaction (SSI) effects in the modelling and analysis procedures, used in the dynamic response analyses. The objective of the project was to evaluate the blast response of the WWER-440/213 Main Building Complex at Paks NPP, based on the data available for the soil properties, recorded free-field blast input motion, and structural design. The scope of EQE-Bulgaria study was to conduct a state-of-the-art SSI analysis with a multiple foundations supported model of the Main Building Complex to assess the structure blast response. The analysis was focused on a modelling technique that assess realistically the SSI effects on the dynamic response of a structure supported on multiple foundation instead of simplified, but more conservative techniques. The scope of research was covered splitting the study into the following steps: development of a twin units model for Main Building Complex structure; development of a Low Strain Soil Properties Model; development of SSI Parameters consisting of a Multiple Foundations System

  13. Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo S, L F

    1991-12-31

    The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author).

  14. Tests of the Bayesian evaluation of SPRT outcomes on Paks NPP data

    International Nuclear Information System (INIS)

    Kulacsy, K.

    1997-01-01

    Self-powered neutron detector signals measured at the Paks Nuclear Power Plant, Hungary, are processed by a simple Binary Sequential Probability Ratio (SPRT) test and an SPRT combined with the Bayesian probability updating method. Since this latter is too robust against changes in the character of the signal, a new version of the Bayesian probability updating method is suggested and tested on the signals. The new method is found to detect signal failures faster and more effectively than the simple SPRT. (R.P.)

  15. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

    2016-09-15

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  16. Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

    International Nuclear Information System (INIS)

    Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

    2013-01-01

    The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

  17. Investigation of radioisotopes in different organisms around Paks NPP

    International Nuclear Information System (INIS)

    Janovics, R.; Bihari, A.; Major, Z.; Palcsu, L.; Papp, L.; Veres, M.

    2010-01-01

    Complete text of publication follows. The Paks Nuclear Power Plant is a pressurised water reactor, therefore, it requires a large amount of cooling water. Cooling water is pumped from the Danube, and used water is also discharged back to the river through the warm-water channel. In the study Danube water and various aquatic organisms (a snail, Viviparus Acerosus, a mussel Unio Tumidus, a predatory fish Stizostedion lucioperca and a non-predatory fish Leuciscus idus) were collected upstream and downstream of the inlet of the channel. After freeze-drying both from the interstitial water and the dry matter of the aquatic organisms collected, tritium measurements were performed by the T- 3 He method to gain information about the ratio of the tritium concentration of the organically bound and the not-bound hydrogen, as well. The activity of the organically bound tritium reflects the mean activity of the environment of the organism, while the tritium activity of the interstitial water shows the actual activity of the aquatic environment. The activity of gamma emitters in the dry matter was also measured by gamma spectrometry. In case of the mussel and snail samples gamma spectrometry measurements were performed separately from the calciferous skeleton and the tissues. Besides the aquatic organisms, soil and plant samples (Scots Pine Pinus sylvestris, Common Milkweed Asclepias syriaca L., giant goldenrod Solidago gigantea) were collected in the vicinity of the nuclear power plant and in a background site, as well. These samples were analysed by gamma spectrometry and for tritium concentration, and the results were compared with a background site. On the basis of the gamma spectrometry results significant amount of artificial gamma emitter isotopes do not get to the Danube through the warm-water channel. Only 60 Co occurred in certain mussel, snail and sludge in a measurable activity concentration, however, it is not of power plant origin, as it was present even in the

  18. Radioactive waste management at WWER type reactors

    International Nuclear Information System (INIS)

    1993-05-01

    This report was prepared within the framework of the Technical Assistance Regional Project on Advice on Waste Management at WWER Type Reactors, which was initiated by the IAEA in 1991. The Regional Project is an integral part of the IAEA's activities directed towards improvement of the safety and reliability of nuclear power plants with WWER type reactors (Soviet designed PWRs). Forty-five WWER type units are currently in operation and twenty-five are under construction in Bulgaria, Czechoslovakia, Finland, Hungary and the former USSR. The idea of regional collaboration between eastern European countries under the auspices of the IAEA was discussed for the first time during the last meeting of the Council for Mutual Economic Assistance (CMEA) on spent fuel and radioactive waste management, held in Rez, Czechoslovakia, in October 1990. Since then, the CMEA and some of its former Member States have ceased to exist. However, there are many reasons for eastern European countries to continue their regional collaboration at a higher level. The USSR, the designer and supplier of WWER type reactors in eastern European countries, participated in the first phase of the project. The majority of WWER type reactors are situated in States of the former USSR (Russia and Ukraine). The main results of the first phase of the Regional Project are: (i) Re-establishment of communication channels among eastern European countries operating WWER type reactors by incorporating the IAEA's technical assistance; (ii) Identification of common waste management problems (administrative and technical) requiring resolution; (iii) Familiarization with radioactive waste management systems at nuclear power plants with WWER type reactors - Paks (Hungary), Loviisa (Finland), Jaslovske Bohunice (Czechoslovakia) and Novovoronezh (Russian Federation). Tabs

  19. D-3He fuel cycles for neutron lean reactors

    International Nuclear Information System (INIS)

    Kernbichler, W.; Miley, G.H.; Heindler, M.

    1989-01-01

    The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

  20. GIT1/beta PIX signaling proteins and PAK1 kinase regulate microtubule nucleation

    Czech Academy of Sciences Publication Activity Database

    Černohorská, Markéta; Sulimenko, Vadym; Hájková, Zuzana; Sulimenko, Tetyana; Sládková, Vladimíra; Vinopal, Stanislav; Dráberová, Eduarda; Dráber, Pavel

    2016-01-01

    Roč. 1863, č. 6 (2016), s. 1282-1297 ISSN 0167-4889 R&D Projects: GA ČR GAP302/12/1673; GA ČR GA15-22194S; GA MŠk LH12050; GA MZd NT14467; GA ČR GA16-23702S Institutional support: RVO:68378050 Keywords : Centrosome * Microtubule nucleation * gamma-tubulin * GIT1/beta PIX signaling proteins * PAK1 kinase Subject RIV: EB - Genetics ; Molecular Biology Impact factor: 4.521, year: 2016

  1. Establishment of a computerized occupational health system at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Otos, M.

    1990-01-01

    An overall personnel health system has been under installation in the Paks Nuclear Power Plant Company, Hungary, for an automatic health monitoring and survey of the personnel exposed to radiation. The system will consist of nine modules when completed. The personnel fitness module is described in detail, and the periodical fitness examinations with computer control are presented. The examinations are using a personnel database system, and a statistical module is used to evaluate monitoring results. (R.P.)

  2. Conceptual design of D-3He FRC reactor 'ARTEMIS'

    International Nuclear Information System (INIS)

    Momota, H.; Ishida, A.; Kohzaki, Y.

    1991-07-01

    A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

  3. The common project for completion of Bubbler Condenser Qualification (Bohunice, Mochovce, Dukovany and Paks NPPs)

    International Nuclear Information System (INIS)

    Jaroslav, H.; Pavol, B.

    2003-01-01

    Described is the common project for completion of bubbler condenser qualification for nuclear power plants in Bohunice, Mochovice, Dukovany and Paks. Functionality of the bubbler condenser was elaborated during the simulation of the main steam line brake, medium break and small break LOCA. On this basis the appropriate operation of bubbler condenser containment under accident conditions can be positively confirmed

  4. ISOLAMENTO TÉRMICO DE RESIDÊNCIAS ATRAVÉS DA REUTILIZAÇÃO DE EMBALAGENS TETRA PAK

    Directory of Open Access Journals (Sweden)

    Jaquiel Salvi Fernandes

    2014-09-01

    Full Text Available A reciclagem está presente na atualidade, não apenas pelo aspecto econômico, mas também pela questão ambiental. Não faz sentido jogar junto com o lixo orgânico materiais que possam ser reaproveitados ou transformados. Neste contexto também se encontram as embalagens de leite e/ou suco longa vida (Tetra Pak®, amplamente utilizadas pela população. Tais embalagens têm baixo valor comercial, e sua reciclagem é difícil e de custo muito elevado. Este trabalho de extensão universitária reutilizou estas embalagens, montando painéis com as dimensões do forro de residências selecionadas na cidade de Videira-SC, com o intuito de isolá-las termicamente. As caixinhas Tetra Pak possuem uma face aluminizada, a qual impede que o calor seja transmitido para o interior (ou exterior no caso do inverno da residência pelo processo de radiação, refletindo mais de 95% do calor. Com esta característica a caixa de leite se mostra perfeita para exercer a função de manta térmica, como uma alternativa às mantas convencionais, com a vantagem de ser uma solução ecológica e barata. Após a instalação, as casas que antes não possuíam forro passaram a registrar temperaturas internas menores no verão e maiores no inverno, além da prevenção contra goteiras e respingos. As famílias atendidas expressaram unanimidade de opinião, mostrando-se muito satisfeitas com o ambiente após a instalação, relatando o aumento da temperatura em dias mais frios e sua diminuição em dias mais quentes. Palavras-chave: painéis térmicos, reutilização, embalagens longa vida, extensão universitária. Thermal isolation of residences through reuse of Tetra Pak packaging Abstract: Nowadays recycling is present not only in economic but also in environmental issues. It makes no sense mixing together with organic waste the materials that can be reused or processed. Milk and long life juice cartons (Tetra Pak®, widely used by the population, are also included

  5. Removal in a lump of JRR-3 nuclear reactor

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Suzuki, Masanori; Nagase, Tetsuo; Watanabe, Morinari.

    1989-01-01

    The research reactor JRR-3 in Japan Atomic Energy Research Institute is called 'Home made No.1 reactor' as all except fuel and heavy water as the moderator and coolant were manufactured in Japan. The JRR-3 attained the criticality in 1962, and the cumulative time of operation reached 47135.5 hours, and the cumulative power output reached 419073.5 MWh. It was stopped in 1983. During the period, it was utilized for beam experiment, irradiation of fuel and materials, RI production and others. In order to cope with the expansion of utilization and the advance of utilizing technology of the research reactor, the reconstruction works are in progress, and the criticality of the reconstructed reactor is expected in 1990. On the site where the old reactor is removed, the reactor of different type is installed, and the first large cold neutron source is equipped. In this report, as to the removal of the old reactor proper, the method of working and the results are described. Considering the period of working, the cost and the management of the removed reactor, in the case of the JRR-3, the method of carrying it out in a lump was adopted as the optimum removal method. The plan, procedure and results of the removal working are reported. (K.I.)

  6. Land use changes in Pak Phanang Basin using satellite images and geographic information system

    Directory of Open Access Journals (Sweden)

    Yongchalermchai, C.

    2004-01-01

    Full Text Available This study defined major changes in land use in Pak Phanang Basin, Nakhon Si Thammarat Province by using remote sensing and geographic information system techniques. The land use map conducted by Department of Land Development in 1988 was compared with the land use map interpreted from satelliteimages of Landsat-5 TM acquired in 1995 and 1999. The results revealed that between 1988 to 1999, forest area in the basin decreased by a total of 98.08 km2, a drastic decline of 60% that was changed to rubber plantation area. The rubber area increased about 181.7 km2 or 41%. Shrimp farm area increased by 184.87 km2, equivalent to a high increase of 886% while paddy field area decreased by 248.7 km2, or 16% that was converted to shrimp farm and rubber land. A decline in forest area caused soil erosion. The severe expansion of shrimp farm area caused the salinity and affected nearby paddy field and water source areas, that resulted in degradation of the environment. Application of remote sensing and geographic information system was utilized as a tool for monitoring the land use change and planning proper resource utilization for sustainable development in Pak Phanang Basin.

  7. 3D simulation of CANDU reactor regulating system

    International Nuclear Information System (INIS)

    Venescu, B.; Zevedei, D.; Jurian, M.

    2013-01-01

    Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

  8. Degradation Behavior and Accelerated Weathering of Composite Boards Produced from Waste Tetra Pak® Packaging Materials

    Science.gov (United States)

    Nural Yilgor; Coskun Kose; Evren Terzi; Aysel Kanturk Figen; Rebecca Ibach; S. Nami Kartal; Sabriye Piskin

    2014-01-01

    Manufacturing panels from Tetra Pak® (TP) packaging material might be an alternative to conventional wood-based panels. This study evaluated some chemical and physical properties as well as biological, weathering, and fire performance of panels with and without zinc borate (ZnB) by using shredded TP packaging cartons. Such packaging material, a worldwide well-known...

  9. Analyses for MARIA Research Reactor with RELAP/MOD3 code

    International Nuclear Information System (INIS)

    Szczurek, J.; Czerski, P.

    2004-01-01

    This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

  10. Completion of reconstruction for Japan Research Reactor No.3

    International Nuclear Information System (INIS)

    Kakefuda, K.; Tani, M.; Isshiki, M.

    1992-01-01

    The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

  11. A facile and rapid automated synthesis of 3'-deoxy-3'-[18F]fluorothymidine

    International Nuclear Information System (INIS)

    Tang Ganghua; Tang Xiaolan; Wen Fuhua; Wang Mingfang; Li Baoyuan

    2010-01-01

    Aim: To develop a simplified and fully automated synthesis procedure of 3'-deoxy-3'-[ 18 F]fluorothymidine ([ 18 F]FLT) using PET-MF-2V-IT-I synthesis module. Methods: Synthesis of [ 18 F]FLT was performed using PET-MF-2V-IT-I synthesis module by one-pot two-step reaction procedure, including nucleophilic fluorination of 3-N-t-butoxycarbonyl-1-[5'-O-(4,4'-dimethoxy triphenylmethyl)-2'-deoxy-3'-O-(4-nitrobenzenesulfonyl) -β-D-threopentofuranosyl]thymine (15 mg) as the precursor molecule with [ 18 F]fluoride, and subsequent hydrolysis of the protecting group with 1.0 M HCl at the same reaction vessel and purification with SEP PAK cartridges instead of the HPLC system. Results: The automated synthesis of [ 18 F]FLT with SEP PAK purification gave corrected radiochemical yield of 23.2±2.6% (n=6, uncorrected yield: 16-22%) and radiochemical purity of >97% within the total synthesis time of 35 min. Conclusion: The fully one-pot automated synthesis procedure with SEP PAK purification can be applied to the fully automated synthesis of [ 18 F]FLT using commercial [ 18 F]FDG synthesis module.

  12. Review of studies pertaining to the seismic input at Paks NPP

    International Nuclear Information System (INIS)

    Muzzi, F.

    1995-01-01

    This report refers to the examination performed on the available material relevant for the seismic input estimate for the Paks NPP, within the frame of the IAEA benchmark study for the seismic analysis and testing of the existing NPPs. The aim of the report is to provide an expert judgement about the quantity and quality of the data and studies performed. The first chapter describes the sources of the data set examined, the second involves the criteria followed in the judgment. The third chapter contains the detailed opinion on the content of the data set, the conclusion and suggestions are reported in chapter four

  13. Time-integrated thyroid dose for accidental releases from Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Raza, S Shoaib; Iqbal, M; Salahuddin, A; Avila, R; Pervez, S

    2004-01-01

    The two-hourly time-integrated thyroid dose due to radio-iodines released to the atmosphere through the exhaust stack of Pakistan Research Reactor-1 (PARR-1), under accident conditions, has been calculated. A computer program, PAKRAD (which was developed under an IAEA research grant, PAK/RCA/8990), was used for the dose calculations. The sensitivity of the dose results to different exhaust flow rates and atmospheric stability classes was studied. The effect of assuming a constant activity concentration (as a function of time) within the containment air volume and an exponentially decreasing air concentration on the time-integrated dose was also studied for various flow rates (1000-50,000 m 3 h -1 ). The comparison indicated that the results were insensitive to the containment air exhaust rates up to or below 2000 m 3 h -1 , when the prediction with the constant activity concentration assumption was compared to an exponentially decreasing activity concentration model. The results also indicated that the plume touchdown distance increases with increasing atmospheric stability. (note)

  14. EL3 reactor description and safety analysis report

    International Nuclear Information System (INIS)

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10 14 neutrons/cm 2 /sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements [fr

  15. Course of operators of the RA-3 reactor

    International Nuclear Information System (INIS)

    Caligiuri, G.A.

    1983-01-01

    Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

  16. New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2017-07-15

    The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

  17. Decontamination of PAH polluted soils by fungi. Subproject: PAH degradation balance and testing of the extended laboratory process. Final report; Dekontamination von PAK belasteten Boeden durch Pilze. Teilprojekt: Bilanzierung des PAK-Abbaus und Erprobung des erweiterten Laborverfahrens. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Martens, R.; Zadrazil, F.; Wolter, M.; Bahadir, M.

    1997-09-01

    The aim of the research project was first to select a fungus with a high potential for mineralization of polycyclic aromatic hydrocarbons (PAH) and a good ability to colonize different soils. The application of this fungus for a degradation of PAH in soil had to be tested. In a screening of 57 white rot fungi the fungus Pleurotus sp. Florida fulfilled these requirements best. In pure culture it was able to metabolize and mineralize highly condensed 4-6 ring PAH to a great extent. For instance, up to 50% of {sup 14}C-pyrene or 39% of {sup 14}C-benzo(a)pyrene was mineralized to {sup 14}CO{sub 2} within 15 weeks. If different carriers for {sup 14}C-pyrene were used the mineralization correlated with the bioavailability, which was characterized by the desorption of the compound from the carriers with water. The mineralization of {sup 14}C-pyrene, {sup 14}C-benz(a)anthracene; {sup 14}C-benzo(a)-pyrene and {sup 14}C-dibenz(a, h)anthracene in native soils showed that a colonization with Pl. sp Florida inhibited the degradation of the less recalcitrant {sup 14}C-pyrene by the indigenous soil microflora. However, the mineralization of the carcinogenic, very recalcitrant and high condensed {sup 14}C-PAH was considerably supported by the fungus. Therefore this capabilities of the fungus could not be proven in a joint medium-scale soil experiment (0.8 m{sup 3} soil) which had been conducted within a parmership with scientists in Jena and an industriell firm. Because of safety aspects only the low condensed less recalcitrant PAH could be applied in this experiment. (orig./MG) [Deutsch] Ziel der Untersuchungen war es, zunaechst aus einer groesseren Zahl von Weissfaeulepilzen Pilze zu selektieren, die ein hohes Abbaupotential fuer PAK besitzen. Fuer die effektive Bildung der fuer den Xenobiotika-Abbau wahrscheinlich verantwortlichen lignolytischen Enzyme sollten die Pilze auf Stroh mit einer Kontamination von {sup 14}C-Pyren angezogen werden. An Hand der Freisetzung von {sup 14

  18. WWER-440 reactor thermal power increase. Up-to-date approaches to substantiation of the core heat-engineering reliability

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Lushin, V.; Zubtsov, D.

    2006-01-01

    Increasing the Units power is an urgent problem for nuclear power plants with WWER-440 reactors. Improving the fuel assembly designs and calculated codes creates all prerequisites to fulfil this purpose. The decrease in the core power peaking is reached by using the profiled fuel assemblies, burnable absorber integrated into the fuel, the FA with the modernized interface attachment, modern calculated codes that allows to reduce conservatism of the RP safety substantiation. A wide spectrum of experimental study of behaviour of the fuel having reached burn-up (50-60) MW days / kg U under the transients and accident conditions was carried out, the post-irradiated examination of the fuel assemblies, fuel rods and fuel pellets with four and five annual operating fuel cycle were performed as well and confirmed the high reliability of the fuel, the presence of large margins of the fuel stack state that contributes to reactor operation at the increased power. The results of the carried out experiments on implementing the five and six annual fuel cycles show that the limiting fuel state as to its serviceability in the WWER-440 reactors is far from being reached. Presently there is an experience of the increased power operation of Kola NPP, Units 1, 2, 4 and Rovno NPP, Unit 2. The Loviisa NPP Units are operated at 109 % power. The Russian experts had gained an experience in substantiating the core operation at 108 % power for Paks NPP, Unit 4. In this paper the additional conditions for increasing the power of the Kola NPP, Units 1 and 2 and the main results of substantiation of increase in power of the Paks NPP, Unit 4 up to 1485 MW are presented in details

  19. “Liting it up”: Popular Culture, Indo-Pak Basketball, and South Asian American Institutions

    OpenAIRE

    Stanley Ilango Thangaraj

    2010-01-01

    South Asian American participants of a co-ethnic basketball league, known as Indo-Pak Basketball, utilized urban basketball vernacular through the phrase “liting it up” to identify individuals scoring points in great numbers. The person “liting it up” becomes visible and receives recognition. Accordingly, I want to “lite up” the scholarship on South Asian America whereby situating South Asian American religious sites and cultural centers as key arenas for “Americanization” through US popula...

  20. Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

    International Nuclear Information System (INIS)

    Boros, I.; Aszodi, A.

    2003-01-01

    The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

  1. Test and evaluation report for Lockheed Idaho Technologies Company, arrow-pak packaging, docket 95-40-7A, type A container

    International Nuclear Information System (INIS)

    Kelly, D.L.

    1996-01-01

    This report incorporates the U.S. Department of Energy, Office of Facility Safety Analysis (DOE/EH-32) approval letter for packaging use. This report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the Arrow-Pak packaging. The Arrow-Pak packaging system consists of Marlex M-8000 Driscopipe, manufactured by Phillips-Driscopipe, Inc., and is sealed with two dome-shaped end caps manufactured from the same materials. The patented sealing process involves the use of electrical energy to heat opposing faces of the pipe and end caps, and hydraulic rams to press the heated surfaces together. This fusion process produces a homogeneous bonding of the end cap to the pipe. The packaging may be used with or without the two internal plywood spacers. This packaging configuration described in this report is designed to ship Type A quantities of solid radioactive materials

  2. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Shoai Tehrani, Bianka; Da Costa, Pascal

    2013-01-01

    Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

  3. The SAFR liquid metal concept

    International Nuclear Information System (INIS)

    Baumeister, E.B.

    1987-01-01

    The Sodium Advanced Fast Reactor (SAFR) modular reactor concept is being developed by the team of Rockwell International, Combustion Engineering, and Bechtel under the U.S. Department of Energy's (DOE's) Advanced Liquid Metal Reactor (LMR) program. The SAFR plant would provide a viable alternate to light water reactors, especially for applications favoring small incremental capacity additions. SAFR is also a logical step to facilitate the later transition to LMFBRs. The SAFR plant concept employs multiple 350-MWe LMR Power Pak modules. Each Power Pak is a standardized, shop-fabricated unit that can be barge-shipped to the plant site for installation. The 350-MWe size allows SAFR to capitalize on all the inherent safety features provided by small reactors and factory fabrication, while still preserving some economy of scale. Shop fabrication minimizes nuclear-grade field fabrication and minimizes the overall plant construction schedule and capital cost. Each Power Pak consists of one reactor assembly and associated heat transfer equipment coupled to a single turbine generator. The reactor core employs mixed uranium-plutonium zirconium alloy metal fuel. The metal-alloy fuel (which has been used in EBR-II) has cost, safety, and safeguard advantages. The intrinsic properties of the sodium coolant (e.g., high boiling point, low vapor pressure, and strong natural convection), blended together with the pool-type LMR concept and the metal fuel, result in an inherently safe plant. Passive inherent features provide both public safety and plant investment protection. Refueling is carried out annually on each Power Pak, replacing one-fourth of the core over a 6-day refueling outage. A colocated pyroprocessing fuel cycle facility can be accommodated at the site such that no off-site shipments are required. (J.P.N.)

  4. D-3He fueled FRC reactor 'ARTEMIS-L'

    International Nuclear Information System (INIS)

    Momota, Hiromu; Tomita, Yukihiro; Ishida, Akio; Kohzaki, Yasuji; Nakao, Yasuyuki; Nishikawa, Masabumi; Ohi, Shoichi; Ohnishi, Masami.

    1992-09-01

    A neutron-lean D- 3 He fueled field reversed configuration (FRC) fusion reactor is studied on the bases of former high-efficiency ARTEMIS design. Certain improvements such as effective axial contracting plasma heating and cusp-type direct energy converters as well as an empirical scale of the energy confinement are introduced. The resultant total neutron load onto the first wall of the plasma chamber is as low as 0.1 MW/m 2 , which enable the life of the first wall or the structural materials to be longer than the whole life of the reactor. The attractive characteristics of the neutron-lean reactor follow in the ARTEMIS design: it is socially acceptable in views of radioactivity and fuel resources, and the cost of electricity appears to be cheap compared with that from a light water reactor. Critical physics and engineering issues for performing the ARTEMIS-L reactor are clarified. (author)

  5. Description of the RA-3 research reactor as a model facility

    International Nuclear Information System (INIS)

    Vicens, Hugo E.; Quintana, Jorge A.

    2001-01-01

    The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

  6. Análise de crescimento e produtividade do pak choi cultivado sob diferentes doses de nitrogênio

    Directory of Open Access Journals (Sweden)

    Janaína Dartora

    2013-08-01

    Full Text Available O objetivo deste trabalho foi avaliar a influência de diferentes doses de nitrogênio no crescimento e produtividade do pak choi. O experimento foi conduzido, em cultivo protegido, de outubro a novembro de 2007, em Marechal Cândido Rondon, PR. O delineamento experimental foi de blocos casualizados, com cinco tratamentos (0, 60, 105, 150 e 195 kg ha-1 de N e quatro repetições. O nitrogênio foi aplicado em três diferentes épocas (transplantio, 7 e 14 dias após o transplantio. Foram realizadas cinco coletas das plantas, semanalmente, avaliando-se a produção de massa da matéria seca e área foliar, para obtenção das taxas de crescimento absoluto e relativo, taxa assimilatória líquida, razão de área foliar e área foliar específica. Na colheita, foram avaliados altura da planta, diâmetro e matéria fresca da parte aérea e produtividade. Incrementos na adubação nitrogenada até a dose de 195 kg ha-1 proporcionam incrementos no crescimento e produtividade do pak choi.

  7. The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

    International Nuclear Information System (INIS)

    Duerigen, S.; Fridman, E.

    2011-01-01

    DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

  8. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

    Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

    1999-01-01

    Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

  9. Experiment for search for sterile neutrino at SM-3 reactor

    Science.gov (United States)

    Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2016-11-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

  10. Integral test of JENDL-3.3 on fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou; Hazama, Taira

    2003-05-01

    An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

  11. Development of an X Window based operator's interface for a core monitoring system

    International Nuclear Information System (INIS)

    Vegh, J.; Huszar, J.; Laz, J.

    1992-09-01

    The components, functioning and programming concepts of the man-machine interface applied in an upgraded version of the core monitoring system and reactor information system VERONA for WWER-440 type nuclear power reactors, installed at the Paks Nuclear Power Plant, are described. The application of the X Window standard Graphical User Interface facilitated modular interface design and made program development easier and faster. (author) 3 refs.; 13 figs

  12. Rac1 is crucial for Ras-dependent skin tumor formation by controlling Pak1-Mek-Erk hyperactivation and hyperproliferation in vivo

    DEFF Research Database (Denmark)

    Wang, Z; Pedersen, Esben Ditlev Kølle; Basse, A

    2010-01-01

    that Rac1 is essential for DMBA/TPA-induced skin tumor formation. This corresponded to a decreased keratinocyte hyperproliferation, although apoptosis was not detectably altered. Activated Rac1 promoted Erk-dependent hyperproliferation by Pak1-mediated Mek activation independent of Mek1 phosporylation...

  13. Ecotoxicological and human toxicological risk assessment of PAH-contaminated soils before and after biological treatment; Oekotoxikologische und humantoxikologische Risikobewertung PAK-belasteter Boeden vor und nach biologischer Behandlung

    Energy Technology Data Exchange (ETDEWEB)

    Roos, P.H.; Hanstein, W.G. [Bochum Univ. (Germany). Inst. fuer Physiologische Chemie; Weissenfels, W.D. [RAG Umwelt Kommunal GmbH, Bottrop (Germany); Afferden, M. van [IMTA, Jiutepec, Mor. (Mexico); Pfeifer, F. [DMT-Gesellschaft fuer Forschung und Pruefung mbH, Essen (Germany)

    2000-07-01

    The goal of the present work is to assess the adverse effects of soil bound polycyclic aromatic hydrocarbons (PAH) which remain in soils after biological remediation. We focus on risk assessment for mammalian species with respect to the oral uptake of contaminated soil particles and compare the results of a biomarker test with those of an ecotoxicological assay, the bioluminescence inhibition test with Vibrio fischeri. As a biomarker effect in mammals, we determined the liver microsomal cytochrome P450 enzyme CYP1A1 which is induced by PAH in exposed rats. After biological soil treatment, different amounts of PAH remain in the soil depending on the soil properties and initial pollutant composition. Particularly, higher condensated PAH resists biological treatment due to its hydrophobicity. In addition, high amounts of organic carbon in the soils affect remediation efficiency. In the bioluminescence inhibition test, eluates of all biologically treated soils studied do not reveal any or only low inhibitory effects. In contrast, the oral uptake of biologically treated contaminated soils leads to induction levels for CYP1A1 similar to those in the untreated samples. A good correlation is obtained between CYP1A1 levels and the amount of 5 and 6-ring PAH in the soil samples. The main result is that the remediation efficiency determined by the luminescence test is not reflected by the biomarker test, a finding which indicates the high bioavailability of residual PAH in soils. Consequently, new criteria for human risk assessment can be delineated. (orig.) [German] Ziel dieser Arbeit ist es, moegliche toxische Wirkungen PAK-belasteter Boeden vor und nach biologischer Sanierung zu erfassen. Hierbei liegt der Schwerpunkt auf der Abschaetzung des Risikos fuer Saeugetiere nach oraler Aufnahme von Bodenpartikeln. Als Biomarker-Effekt fuer die PAK-Aufnahme haben wir in Ratten die Induktion des lebermikrosomalen P450-Enzyms CYP1A1 bestimmt, dessen Expression durch PAK moduliert

  14. 3D CAD model of the subcritical nuclear reactor of IPN

    International Nuclear Information System (INIS)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

    2016-09-01

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  15. Increased Circulating Endothelial Microparticles Associated with PAK4 Play a Key Role in Ventilation-Induced Lung Injury Process

    Directory of Open Access Journals (Sweden)

    Shuming Pan

    2017-01-01

    Full Text Available Inappropriate mechanical ventilation (MV can result in ventilator-induced lung injury (VILI. Probing mechanisms of VILI and searching for effective methods are current areas of research focus on VILI. The present study aimed to probe into mechanisms of endothelial microparticles (EMPs in VILI and the protective effects of Tetramethylpyrazine (TMP against VILI. In this study, C57BL/6 and TLR4KO mouse MV models were used to explore the function of EMPs associated with p21 activated kinases-4 (PAK-4 in VILI. Both the C57BL/6 and TLR4 KO groups were subdivided into a mechanical ventilation (MV group, a TMP + MV group, and a control group. After four hours of high tidal volume (20 ml/kg MV, the degree of lung injury and the protective effects of TMP were assessed. VILI inhibited the cytoskeleton-regulating protein of PAK4 and was accompanied by an increased circulating EMP level. The intercellular junction protein of β-catenin was also decreased accompanied by a thickening alveolar wall, increased lung W/D values, and neutrophil infiltration. TMP alleviated VILI via decreasing circulating EMPs, stabilizing intercellular junctions, and alleviating neutrophil infiltration.

  16. Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

    International Nuclear Information System (INIS)

    Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

    2009-01-01

    The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

  17. Estimation of reactor pool water temperature after shutdown in JRR-3M

    International Nuclear Information System (INIS)

    Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

    1999-01-01

    The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

  18. RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Debrecin, N.; Foretic, D.

    2003-01-01

    In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

  19. Modelling of MOCVD Reactor: New 3D Approach

    Science.gov (United States)

    Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

    2014-04-01

    The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

  20. Modelling of MOCVD reactor: new 3D approach

    International Nuclear Information System (INIS)

    Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

    2014-01-01

    The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

  1. Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

    Science.gov (United States)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

  2. Experimental investigation and numerical modelling of tritium wash-out by precipitation in the area of the nuclear power plant of Paks, Hungary

    International Nuclear Information System (INIS)

    Koelloe, Z.; Palcsu, L.; Major, Z.; Papp, L.; Molnar, M.

    2009-01-01

    Complete text of publication follows. Tritium is an important radioactive isotope because its natural occurrence in the air and precipitation due to natural and artificial sources. In order to investigate the natural changes in tritium concentration, the artificial component has to be known. Some field experiments were carried out before to investigate the washout of tritium by precipitation emitted from artificial sources, but none of them were carried out around a real power plant, measuring the deposition pattern. We collected rainwater around the nuclear power plant of Paks (Paks NPP), and analyzed them for tritium. The rainwater samplers were constructed at ATOMKI. Their structure consists of a funnel, support parts and a long tube, acting as the storage vessel for the rainwater. One end of the tube is connected to the funnel, the other end is open. This way when 'new' rainwater falls, it pushes out the 'old' water, causing that always the last rainwater is in the tube (if enough rain is falling). In total 54 samplers were placed out around the Paks NPP in two half circles, with radiuses 400 and 800 m, pointing east. We collected samples after a rain period on 7-8 June 2009. They were prepared and measured with liquid scintillation counting (LSC) for tritium. We measured some samples also with the 3 He-ingrowth method, to ensure better accuracy. The measurement results in Fig. 1 clearly show the trace of the tritium plume emitted from the plant. However, the highest values are not very high, compared to environmental levels, and considering the fact that all the samples were collected from the area of the plant. A numerical model was coded to calculate the washout of tritium theoretically from the meteorological and emission data, and to estimate the effect of the plant in larger distances. In Fig. 1 it is apparent that the model, however does not describe the data precisely, but gives reasonable results, especially for the outer circle. The calculations also

  3. PR-EDB: Power Reactor Embrittlement Database Version 3

    International Nuclear Information System (INIS)

    Wang, Jy-An John; Subramani, Ranjit

    2008-01-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  4. PR-EDB: Power Reactor Embrittlement Database - Version 3

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  5. Influence of tensides and lipophilic substrates on the biological availability of polycyclic aromatic hydrocarbons (PAHs); Ueber dem Einfluss von Tensiden und lipophilen Substraten auf die Bioverfuegbarkeit von polyzyklischen aromatischen Kohlenwasserstoffen (PAK)

    Energy Technology Data Exchange (ETDEWEB)

    Soeder, C.J. von; Kleespies, M; Eschner, C; Webb, L; Groeneweg, J [Forschungszentrum Juelich GmbH (Germany). IBT-3/ICG-6

    1998-12-31

    The objects of the study were as follows: isolation and characterization of PAH-degrading micro-organisms from lysimeters; tests relating to the experimental simulation of the conditions permitting pollutant degradation in soil; investigation of the influence of tensides and other dissolved organic compounds on the biological availability and degradation of PAHs. (orig./SR) [Deutsch] - Isolierung und Charakterisierung PAK-abbauender Mikroorganismen aus Lysimetern; Versuche zur experimentellen Simulation der Bedingungen, unter denen der Abbau von Schadstoffen im Boden erfolgt. - Untersuchung des Einflusses von Tensiden und anderen geloesten organischen Verbindungen auf Bioverfuegbarkeit und Abbau von PAK. (orig./SR)

  6. Influence of tensides and lipophilic substrates on the biological availability of polycyclic aromatic hydrocarbons (PAHs); Ueber dem Einfluss von Tensiden und lipophilen Substraten auf die Bioverfuegbarkeit von polyzyklischen aromatischen Kohlenwasserstoffen (PAK)

    Energy Technology Data Exchange (ETDEWEB)

    Soeder, C.J. von; Kleespies, M.; Eschner, C.; Webb, L.; Groeneweg, J. [Forschungszentrum Juelich GmbH (Germany). IBT-3/ICG-6

    1997-12-31

    The objects of the study were as follows: isolation and characterization of PAH-degrading micro-organisms from lysimeters; tests relating to the experimental simulation of the conditions permitting pollutant degradation in soil; investigation of the influence of tensides and other dissolved organic compounds on the biological availability and degradation of PAHs. (orig./SR) [Deutsch] - Isolierung und Charakterisierung PAK-abbauender Mikroorganismen aus Lysimetern; Versuche zur experimentellen Simulation der Bedingungen, unter denen der Abbau von Schadstoffen im Boden erfolgt. - Untersuchung des Einflusses von Tensiden und anderen geloesten organischen Verbindungen auf Bioverfuegbarkeit und Abbau von PAK. (orig./SR)

  7. Operating reactors licensing actions summary. Vol. 3, No. 3

    International Nuclear Information System (INIS)

    1983-04-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  9. Automatic acoustic and vibration monitoring system for nuclear power plants

    International Nuclear Information System (INIS)

    Tothmatyas, Istvan; Illenyi, Andras; Kiss, Jozsef; Komaromi, Tibor; Nagy, Istvan; Olchvary, Geza

    1990-01-01

    A diagnostic system for nuclear power plant monitoring is described. Acoustic and vibration diagnostics can be applied to monitor various reactor components and auxiliary equipment including primary circuit machinery, leak detection, integrity of reactor vessel, loose parts monitoring. A noise diagnostic system has been developed for the Paks Nuclear Power Plant, to supervise the vibration state of primary circuit machinery. An automatic data acquisition and processing system is described for digitalizing and analysing diagnostic signals. (R.P.) 3 figs

  10. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

  11. Nuclear research reactor 0.5 to 3 MW

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-05-15

    This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

  12. Nuclear research reactor 0.5 to 3 MW

    International Nuclear Information System (INIS)

    1992-05-01

    This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

  13. EL-3 dismantling of an experimental reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

  14. An assessment of the parameters and experimental data of the continuous water activity monitors operating in the upstream and downstream channels of the Paks nuclear power plant

    International Nuclear Information System (INIS)

    Nagy, Gy.; Feher, I.

    1986-03-01

    A NaI(Tl) scintillator was placed into a measuring vessel of 8 msup(3) volume for monitoring the effluents in the upstream and downstream channels of the Paks nuclear power plant. The effects of radioactivity, meteorological parameters, and the atmospheric pressure on the counting rates, and their daily and monthly average values in both channels were analyzed. The short-term increases of the monitor signals could be attributed to rainy weather. The sup(222)Rn countent of water was also evaluated. (author)

  15. Automatic closed-loop stereo-photogrammetric system for the Nuclear Power Station Paks

    International Nuclear Information System (INIS)

    Eoery, K.; Szabados, J.; Szerovay, A.

    1982-01-01

    The geodesic work for the NPS Paks project required an extensive modernization of traditional measuring techniques, besides the development of measuring devices and methods encompassing the complete procedure of data processing. Stereo-photogrammetry based on three-dimensional measuring technique plays an outstanding role in 'bulk work' required for measuring as well as in efficient technical data supply. A detailed analysis of the technical parameters is given concerning the interactive graphic and digital data processing and data base organizing system. After the description of the main types of process equipment system organization problems are discussed. The paper outlines research and testing tasks related to the practical application of the technology system of world standard, representing a unique solution in Hungarian relation. Finally, fields of application for the system in power station design are presented. (author)

  16. Periodical safety review of units 1 and 2 of PAKS NPP. Examples from summary report

    International Nuclear Information System (INIS)

    Hammar, K.

    1998-01-01

    On the basis of American practice of qualification and relevant IAEA recommendations detailed guidelines of the qualification procedure were developed and executed on the Units 1 and 2 of the Paks NPP. Periodic safety supervision will be performed by evaluation of the following reports to be submitted by NPP: real technical conditions of the facility; existing practice and proposals for equipment qualification; evaluation of the existing safety reports estimating their validity up to the plant lifetime; ageing and ageing management; procedures of operation, maintenance, supervision; organisation and administration; safety impact of human factor, training, education, qualification of personnel

  17. Optimal organization of structural analysis and site inspection for the seismic requalification of Paks NPP

    International Nuclear Information System (INIS)

    Contri, P.

    1996-01-01

    The analysis described in this report deals with a numerical procedure aimed for the assessment of a methodology for the optimal organization of data collection, in the context of seismic requalification of structures and components of existing nuclear power plants. The presented procedure has quite a general application and an example was chosen for the Paks NPP where seismic requalification is in progress. The assessment was carried out in reference to the following main tasks: structure and soil data analysis; numerical model generation; deterministic dynamic analysis description; reliability analysis framework discussion; transfer function calculation via response surface approach; and the sensitivity evaluation

  18. Use of plate fuel elements for the RA3 reactor

    International Nuclear Information System (INIS)

    Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

    1992-01-01

    The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

  19. Inherent controllability in modular ALMRs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Sevy, R.H.; Wei, T.Y.C.

    1989-01-01

    As part of recent development efforts on advanced reactor designs ANL has proposed the IFR (Integral Fast Reactor) concept. The IFR concept is currently being applied to modular sized reactors which would be built in multiple power paks together with an integrated fuel cycle facility. It has been amply demonstrated that the concept as applied to the modular designs has significant advantages in regard to ATWS transients. Attention is now being focussed on determining whether or not those advantages deriving from the traits of the IFR can be translated to the operational/DBA (design basis accident) class of transients. 5 refs., 3 figs., 3 tabs

  20. Characteristics of D-3He fueled frc reactor: ARTEMIS-L

    International Nuclear Information System (INIS)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

  1. Summary of the 3rd workshop on the reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  2. Summary of the 3rd workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  3. Physical, mechanical and hydration kinetics of particleboards manufactured with woody biomass (Cupressus lusitanica, Gmelina arborea, Tectona grandis), agricultural resources, and Tetra Pak packages.

    Science.gov (United States)

    Moya, Róger; Camacho, Diego; Oporto, Gloria S; Soto, Roy F; Mata, Julio S

    2014-02-01

    Lignocellulosic wastes resulting from agricultural activities as well as Tetra Pak residues from urban centres can cause significant levels of pollution. A possible action to minimize this problem is to use them in the production of particleboards. The purpose of this study was to evaluate the physical, mechanical, and hydration properties of particleboards manufactured with the mixture of woody biomass (Cupressus lusitanica, Gmelina arborea, and Tectona grandis) and either agricultural wastes [pineapple leaves (Ananas comosus) and palm residues (Elaeis guineensis)] or Tetra Pak residues (TP). The results show that the particleboards prepared with TP and woody biomass can reduce the swelling and water absorption in up to 40% and 50% compared with particleboards without TP. Also, these particleboards had increased flexure resistance and shear stress (up to 100%) compared with those without TP. On the contrary, particleboards prepared with pineapple leaves in combination with woody biomass showed the lowest mechanical properties, particularly for tensile strength, hardness, glue-line shear, and nail and screw evaluation.

  4. Autocrine VEGF and IL-8 Promote Migration via Src/Vav2/Rac1/PAK1 Signaling in Human Umbilical Vein Endothelial Cells.

    Science.gov (United States)

    Ju, Li; Zhou, Zhiwen; Jiang, Bo; Lou, Yue; Guo, Xirong

    2017-01-01

    Pro-angiogenic factors VEGF and IL-8 play a major role in modulating the migratory potential of endothelial cells. The goal of this study was to investigate the effect of autocrine VEGF and IL-8 in the form of self-conditioned medium (CM) on human umbilical vein endothelial cells (HUVECs). Enzyme-linked immunosorbent assay (ELISA) examined the automatic secretion of VEGF and IL-8 protein by HUVECs. Western blot, small interfering RNA (siRNA), pulldown and Transwell assays were used to explore the role and the mechanism of autocrine VEGF and IL-8 in migration of HUVECs. Neutralizing VEGF and IL-8 in CM significantly abrogated CM-induced migration of HUVECs. Autocrine VEGF and IL-8 increased Src phosphorylation, Rac1 activity and PAK1 phosphorylation in a time dependent manner. Additionally, blocking Rac1 activity with Rac1 siRNA largely abolished autocrine VEGF and IL-8-induced cell migration. Vav2 siRNA suppressed autocrine VEGF and IL-8-induced Rac1 activation and cell migration. Furthermore, blocking Src signaling with PP2, a specific inhibitor for Src, markedly prevented autocrine VEGF and IL-8-induced Vav2 and Rac1 activation as well as consequently cell migration. PAK1 siRNA also significantly abolished autocrine VEGF and IL-8-induced cell migration. We demonstrated for the first time that autocrine VEGF and IL-8 promoted endothelial cell migration via the Src/Vav2/Rac1/PAK1 signaling pathway. This finding reveals the molecular mechanism in the increase of endothelial cell migration induced by autocrine growth factors and cytokines, which is expected to provide a novel therapeutic target in vascular diseases. © 2017 The Author(s)Published by S. Karger AG, Basel.

  5. Integral test of JENDL-3.3 for thermal reactors

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Mori, Takamasa

    2003-01-01

    Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

  6. Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

    International Nuclear Information System (INIS)

    Shi, Dunfu

    2015-01-01

    The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

  7. Computerized reactor protection and safety related systems in nuclear power plants. Proceedings of a specialists' meeting. Working material

    International Nuclear Information System (INIS)

    1998-01-01

    Though the majority of existing control and protection systems in nuclear power plants use old analogue technology and design philosophy, the use of computers in safety and safety related systems is becoming a current practice. The Specialists Meeting on ''Computerized Reactor Protection and Safety Related Systems in Nuclear Power Plants'' was organized by IAEA (jointly by the Division of Nuclear Power and the Fuel Cycle and the Division of Nuclear Installation Safety), in co-operation with Paks Nuclear Power Plant in Hungary and was held from 27-29 October 1997 in Budapest, Hungary. The meeting focused on computerized safety systems under refurbishment, software reliability issues, licensing experiences and experiences in implemented computerized safety and safety related systems. Within a meeting programme a technical visit to Paks NPP was organized. The objective of the meeting was to provide an international forum for the presentation and discussion on R and D, in-plant experiences in I and C important to safety, backfits and arguments for and reservations against the digital safety systems. The meeting was attended by 70 participants from 16 countries representing NPPs and utility organizations, design/engineering, research and development, and regulatory organizations. In the course of 4 sessions 25 technical presentations were made. The present volume contains the papers presented by national delegates and the conclusions drawn from the final general discussion

  8. Failed (leaking) spent fuel management and storage in the Paks NPP

    International Nuclear Information System (INIS)

    Burjan, T.

    2011-01-01

    At the cycle 22, unit 4, Paks NPP the fissile contents raised irregularly in the water of the primary circuit. At the end of the cycle sipping tests were performed for the entire core to find out the leaking fuel assembly primarily responsible for this phenomenon. The identified leaking assembly temporarily was placed in the Spent Fuel Relaxing Pool. For measuring environmental impact of leaking assemblies an investigation program was developed and implemented. The assessment covered the following: effects of the leaking fuel on the water of relaxing pool and on the gaseous emissions in case open storage; in case when the leaking cassette is in a special hermetical storage case, how much gas is collected in the locked case and what is its composition; how to change the measured sipping test signal depending on relaxing time of leaking fuel cassettes. Based on the evaluation of the investigation program results the NPP modified the operational instructions for the treatment and storage of failed fuel assemblies. (author)

  9. A Trio-Rac1-PAK1 signaling axis drives invadopodia disassembly

    Science.gov (United States)

    Moshfegh, Yasmin; Bravo-Cordero, Jose Javier; Miskolci, Veronika; Condeelis, John; Hodgson, Louis

    2014-01-01

    Rho family GTPases control cell migration and participate in the regulation of cancer metastasis. Invadopodia, associated with invasive tumor cells, are crucial for cellular invasion and metastasis. To study Rac1 GTPase in invadopodia dynamics, we developed a genetically-encoded, single-chain Rac1 Fluorescence Resonance Energy Transfer (FRET) biosensor. The biosensor shows Rac1 activity exclusion from the core of invadopodia, and higher activity when invadopodia disappear, suggesting that reduced Rac1 activity is necessary for their stability, and Rac1 activation is involved in disassembly. Photoactivating Rac1 at invadopodia confirmed this previously-unknown Rac1 function. We built an invadopodia disassembly model, where a signaling axis involving TrioGEF, Rac1, PAK1, and phosphorylation of cortactin, causing invadopodia dissolution. This mechanism is critical for the proper turnover of invasive structures during tumor cell invasion, where a balance of proteolytic activity and locomotory protrusions must be carefully coordinated to achieve a maximally invasive phenotype. PMID:24859002

  10. Primary water chemistry control at units of Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Schunk, J.; Patek, G.; Pinter, T.; Tilky, P.; Doma, A.; Osz, J.

    2010-01-01

    The primary water chemistry of the four identical units of Paks Nuclear Power Plant has been developed based on Western-type PWR units, taking into consideration some Soviet-Russian modifications. The political changes in 90s have also influenced the water chemistry specifications and directions. At PWR units the transition operational modes have been developed while in case of VVER units - in lack of central uniform regulation - this question has become the competence and responsibility of each individual plant. This problem has resulted in separate water chemistry developments with a considerable time delay. The needs for life-time extensions all over the World have made the development of start-up and shut-down chemistry procedures extremely important, since they considerably influence the long term and safe operation of plants. The uniformly structured limit value system, the principles applied for the system development, and the logic schemes for actions to be taken are discussed in the paper, both for normal operation and transition modes. (author)

  11. Primary Water Chemistry Control at Units of Paks Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Schunk, J.; Pinter, G. Patek T.; Tilky, P.; Doma, A. [Paks Nuclear Power Plant Co. Ltd., Paks (Hungary); Osz, J. [Budapest University of Technology and Economics, Budapest (Hungary)

    2013-03-15

    The primary water chemistry of the four identical units of Paks Nuclear Power Plant has been developed based on Western type PWR units, taking into consideration some Russian modifications. The political changes in the 1990s have also influenced the water chemistry specifications and directions. At PWR units the transition operational modes have been developed while in case of WWER units - in lack of central uniform regulation - this question has become the competence and responsibility of each individual plant. This problem has resulted in separate water chemistry developments with a considerable time delay. The need for lifetime extensions worldwide has made the development of startup and shutdown chemistry procedures extremely important, since they considerably influence the long term and safe operation of plants. The uniformly structured limit value system, the principles applied for the system development, and the logic schemes for actions to be taken are discussed in the paper, both for normal operation and transition modes. (author)

  12. Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

    CERN Document Server

    Chiba, G

    2002-01-01

    It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

  13. Development of 3D CFD simulation method in nuclear reactor safety analysis

    International Nuclear Information System (INIS)

    Rosli Darmawan; Mariah Adam

    2012-01-01

    One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

  14. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Takats, F [TS Enercon Kft. (Hungary)

    2012-07-01

    Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. The fresh fuel is imported from Russia so far. The spent fuel assemblies were shipped back to Russia until 1997 after about 6 years cooling at the plant. A dry storage facility (MVDS type) has been constructed and is operational since then. By 1 January 2008, there were 5107 assemblies in dry storage. The objectives are: 1) Wet AR storage of spent fuel from the NPP Paks: Measurements of conditions for spent fuel storage in the at-reactor (AR) storage pools of Paks NPP (physical and chemical characteristics of pool water, corrosion product data); Measurements and visual control of storage pool component characteristics; Evaluation of storage characteristics and conditions with respect to long-term stability (corrosion of fuel cladding, construction materials); 2) Dry AFR storage at Paks NPP: Calculation and measurement of spent fuel conditions during the transfer from the storage pool to the modular vault dry storage (MVDS) on the site; Calculation and measurement of spent fuel conditions during the preparation of fuel for dry storage (drying process), such as crud release, activity build-up; Measurement of spent fuel conditions during the long-term dry storage, activity data in the storage tubes and amount of crud.

  15. Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

    International Nuclear Information System (INIS)

    Tseng, C.M.; Lepp, R.M.

    1983-07-01

    A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

  16. RA reactor exploitation, task 3.08/01

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-01-01

    During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

  17. Analysis of Paks NPP Personnel Activity during Safety Related Event Sequences

    International Nuclear Information System (INIS)

    Bareith, A.; Hollo, Elod; Karsa, Z.; Nagy, S.

    1998-01-01

    Within the AGNES Project (Advanced Generic and New Evaluation of Safety) the Level-1 PSA model of the Paks NPP Unit 3 was developed in form of a detailed event tree/fault tree structure (53 initiating events, 580 event sequences, 6300 basic events are involved). This model gives a good basis for quantitative evaluation of potential consequences of actually occurred safety-related events, i.e. for precursor event studies. To make these studies possible and efficient, the current qualitative event analysis practice should be reviewed and a new additional quantitative analysis procedure and system should be developed and applied. The present paper gives an overview of the method outlined for both qualitative and quantitative analyses of the operator crew activity during off-normal situations. First, the operator performance experienced during past operational events is discussed. Sources of raw information, the qualitative evaluation process, the follow-up actions, as well as the documentation requirements are described. Second, the general concept of the proposed precursor event analysis is described. Types of modeled interactions and the considered performance influences are presented. The quantification of the potential consequences of the identified precursor events is based on the task-oriented, Level-1 PSA model of the plant unit. A precursor analysis system covering the evaluation of operator activities is now under development. Preliminary results gained during a case study evaluation of a past historical event are presented. (authors)

  18. Automation of the electromagnetic filter applied for condensation water treatment in the secondary cooling system of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Szilagyi, Gyoergy

    1989-01-01

    A full-flow condensation water purification system is applied in the secondary cooling circuit of the Paks NPP. The electromagnetic filter of the filtering system eliminates ferromagnetic impurities. The filter consists of a high current coil and an automatic control unit. During the improvement of this unit, a FESTO FPC-404 type controller based on an extended capability PLC was installed. (R.P.) 5 figs

  19. Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

    Energy Technology Data Exchange (ETDEWEB)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

  20. Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

    International Nuclear Information System (INIS)

    Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

    1982-03-01

    A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

  1. Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

    Energy Technology Data Exchange (ETDEWEB)

    Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

    2017-04-01

    Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

  2. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  3. 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.)

  4. Selected examples for safety analysis in VVER-440 type reactors simulated by the coupled ATHLET/KIKO3D code system

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

    2005-01-01

    Recently several projects have been initiated in Hungary aiming at the introduction of new fuel type, increased maximum allowed power and economic fuel cycle. The planned upgraded power and parallel application of new fuel type require the renewal of the relevant chapter of the Final Safety Analysis Report (FSAR). One of the main tools used for analyzing transient scenarios initiating by reactivity and power distribution anomalies was the ATHLET/KIKO3D coupled neutron kinetic / thermal-hydraulic code. This paper gives an overview of two analyses, which was prepared in the frame of the revision of Paks FSAR, namely the ''withdrawal of one control rod'' and ''initial phase of main steam line break'' events. (author)

  5. Ignition access in a D-3He helical reactor

    International Nuclear Information System (INIS)

    Mitarai, Osamu

    2003-01-01

    Ignition access in a D- 3 He helical reactor is studied based on 0-dimensional particle and power balance equations for deuterium, tritium, helium-3, alpha ash, proton ash, electron density and temperature. The calculations are based on the following experimental facts observed in LHD. (author)

  6. Fusion blankets for catalyzed D--D and D--He3 reactors

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

  7. Fusion blankets for catalyzed D--D and D--3He reactors

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

  8. RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

    International Nuclear Information System (INIS)

    Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

    2003-01-01

    This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

  9. UnPAKing RUNX3 functions-Both sides of the coin.

    Science.gov (United States)

    Kumar, Arun; Sundaram, Sandhya; Rayala, Suresh K; Venkatraman, Ganesh

    2017-06-19

    Post translational modifications of RUNX3 have been shown to play an important role in directing RUNX3 functions. In this review we highlight the phosphorylation dependent functions of RUNX3 as regulated by PAK1 and its implications on tumorigenesis.

  10. Design and implementation experience of seismic upgrades at Kozloduy and Paks NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Borov, V; Trichkov, V; Alexandrov, A; Jordanov, M [EQE-Bulgaria, Sofia (Bulgaria)

    1995-07-01

    Series of upgrades have been designed and implemented by EQE-Bulgaria at Kozloduy NPP and as a subcontractor of EQE-International - at Paks NPP. Wide variety of facilities have been upgraded, including Electrical Equipment, Control and Instrumentation Equipment, Technological Equipment, Brick Walls and Building Structures. Different design approaches and concepts have been applied in compliance with the specific technological and structural conditions. The effect of the excitation intensity as well as the presence of specific floor response spectra over the upgrading concept and cost is discussed. Specific problems of supporting heavy technological equipment are noted. A practical approach for seismic upgrading of Brick Walls, as well as a tendency for unification of the engineering design is shown. The first completely upgraded Building Structure at Kozloduy NPP is the structure of the Electrical Control Building to the Diesel Generator of the River-bank Pump Station. Specific problems of the implementation of the final upgrading design of the Diesel Generator Building are outlined. (author)

  11. Design and implementation experience of seismic upgrades at Kozloduy and Paks NPPs

    International Nuclear Information System (INIS)

    Borov, V.; Trichkov, V.; Alexandrov, A.; Jordanov, M.

    1995-01-01

    Series of upgrades have been designed and implemented by EQE-Bulgaria at Kozloduy NPP and as a subcontractor of EQE-International - at Paks NPP. Wide variety of facilities have been upgraded, including Electrical Equipment, Control and Instrumentation Equipment, Technological Equipment, Brick Walls and Building Structures. Different design approaches and concepts have been applied in compliance with the specific technological and structural conditions. The effect of the excitation intensity as well as the presence of specific floor response spectra over the upgrading concept and cost is discussed. Specific problems of supporting heavy technological equipment are noted. A practical approach for seismic upgrading of Brick Walls, as well as a tendency for unification of the engineering design is shown. The first completely upgraded Building Structure at Kozloduy NPP is the structure of the Electrical Control Building to the Diesel Generator of the River-bank Pump Station. Specific problems of the implementation of the final upgrading design of the Diesel Generator Building are outlined. (author)

  12. Inspections of CRDM Nozzle Penetrations in Paks NPP

    International Nuclear Information System (INIS)

    Doszpod, B.; Doczi, M.

    2008-01-01

    During the maintenance outage of Unit 2 of Paks Nuclear Power Plant in 2002, performing the regular drop-test of Control Rod Driving Mechanisms (CRDM) reduced drop-speed was observed in case of one CRDM. In spite of the measured value of speed was inside the acceptance limit, so it was still satisfactory, decision was made to disassemble the CRDM to clarify the cause of the speed-anomaly. After removal of the CRDM, by means of visual inspection deformation (bulge) was observed on the inside surface of the heat protection tube of the CRDM nozzle penetration. Deformation was big enough to obstruct the free movement of CRDM. After the deformed heat protection tube was removed, significant bulge was observed also on the corrosion protection tube, at the same elevation. As the root cause of these deformations, presence of water in the space between the CRDM nozzle and the corrosion protection tube was assumed. Non destructive inspection procedures were worked out and utilized to detect the presence of water in the space in question and to find the possible way of water inlet. Performed inspections successfully localized the place of water inlet. Developed inspection program of CRDM nozzles has to be performed during each outage on each unit. Paper deals with introduction of the phenomenon, the cause of damage, inspection the procedures which were worked out and applied, summarize the results of inspections performed.(author)

  13. Chilean experience in production of 18F-FDG from 18F in a reactor

    International Nuclear Information System (INIS)

    Chandia, M.; Godoy, N.; Errazu, X.; Hernandez; Figols, M.; Firnau, G.; Tronsoco, F.

    2000-01-01

    18 F-FDG (fluorine-deoxy-D-glucose) is an important and useful radiopharmaceutical for imaging and study of myocardial viability. Usually cyclotron-produced 18 F is used to label 18 F-FDG. The availability of a 5 MW Nuclear Reactor in Chile and the absence of a quality cyclotron to produce 18 F required that we developed a method in order to obtain suitable 18 F to label 18 F-FDG using the facilities we have at the Nuclear Center of La Reina, Chilean Nuclear Energy Commission. The nuclear reactions involved are: 6 Li(n,aα) 3 H and 16 O( 3 H,n) 18 F. Enriched Li 2 CO 3 ( 6 Li = 95 %) was irradiated in a 5 MW swimming pool type nuclear reactor with a neutron flux of 5. 7 x 10 13 n cm -2 s -1 for 4 hours. The irradiated Li 2 CO 3 was dissolved in H 2 SO 4 (1:1) and distilled as trimethylsilyl( 18 F)fluoride ( 18 F-TMS). The labelling of the sugar was carried out using the method described by Hamacker. The 18 F-TMS was trapped in a solution of acetonitrile, water, potassium carbonate, and kriptofix and hydrolysed to form 18 F fluoride. The nucleophilic complex reacts with 1,3,4,6, tetra-O-acetyl- 2-O-trifluoromethanesulfonyl-bβ-D-mannopyranose. The acetylated carbohydrate by acid hydrolysis produces 18 F-FDG. The final product was purified using an ion retarding resin (AG11-A8) and a system two Sep Pak Plus: Alumina and C-18 cartridge and sterilised by Millipore 0.22 μm filter. The 18 F-FDG was obtained in an apyrogenic and sterile solution. The 18 F radionuclide purity was higher than 99.9% and the radiochemical purity ofthe 18 F-FDG obtained was over than 99%. Residual 3 H content was as low as 20 (Bq 3 H/MBq 18 F-FDG.). The yield of the process 18 F-FDG was 13.2 %. (authors)

  14. GRIMH3: A new reactor calculation code at Savannah River Site

    International Nuclear Information System (INIS)

    Le, T.T.; Pevey, R.E.

    1993-01-01

    The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

  15. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  16. Rapid purification of radioiodinated peptides with Sep-Pak reversed phase cartridges and HPLC

    International Nuclear Information System (INIS)

    Miller, J.J.; Schultz, G.S.; Levy, R.S.

    1984-01-01

    A simple, rapid method is described for the purification of radioiodinated peptides for use in radioimmuno- and in radioreceptor assays. Iodinated reaction mixtures are applied directly onto Sep-Pak disposable, reversed phase cartridges equilibrated with phosphate buffer. Unreacted 125-iodide and other non-peptide reaction components are eluted with buffer. The peptide fraction is then eluted with 70% buffer:30% acetonitrile. The peptide fraction is further purified by reversed phase high pressure liquid chromatography to separate the native peptide and the mono- and diiodo-derivatives. In this study the method is used to prepare 125-iodide-labeled monoiodo-leucine enkephalin and monoiodo-angiotensin II, which are free of the parent peptides and diiodo-derivatives and are of maximum obtainable specific radioactivity. The usefulness of these labeled peptides in radioimmuno- and radioreceptor assays is demonstrated by their binding to specific antibodies and receptors, respectively. (author)

  17. Leak detection of KNI seals

    International Nuclear Information System (INIS)

    Baranyai, G.; Peter, A.; Windberg, P.

    1990-03-01

    In Unit 3 and 4 of the Paks Nuclear Power Plant, Hungary, KNI type seals are used as lead-throughs with conical nickel sealing rings. Their failure can be critical for the operation of the reactor. An Acoustical Leak Detection System (ALDS) was constructed and tested for the operational testing of the seals. Some individual papers are presented in this collection on the calibration and testing of the ALDS intended to be placed on the top of the reactor vessels. The papers include simulation measurements of Unit 3 of NPP, laboratory experiments, evaluation of measurements, and further development needs with the ALDS. (R.P.) 50 figs.; 19 tabs

  18. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  19. Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

    International Nuclear Information System (INIS)

    Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

    2012-01-01

    10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

  20. Adhesion inhibition of F1C-fimbriated Escherichia coli and Pseudomonas aeruginosa PAK and PAO by multivalent carbohydrate ligands.

    Science.gov (United States)

    Autar, Reshma; Khan, A Salam; Schad, Matthias; Hacker, Jörg; Liskamp, Rob M J; Pieters, Roland J

    2003-12-05

    In order to evaluate their inhibition of bacterial adhesion, the carbohydrate sequences GalNAcbeta1-->4Gal and GalNAcbeta1-->4Galbeta1-->4Glc were synthesized. The disaccharide was conjugated to dendrons based on the 3,5-di-(2-aminoethoxy)-benzoic acid branching unit to yield di- and tetravalent versions of these compounds. A divalent compound was also prepared that had significantly longer spacer arms. Relevant monovalent compounds were prepared for comparison. Their anti-adhesion properties against F1C-fimbriated uropathogenic Escherichia coli were evaluated in an ELISA-type assay by using a recombinant strain and also by using Pseudomonas aeruginosa strains PAO and PAK. Adhesion inhibition was observed in all cases, and multivalency effects of up to one order of magnitude were observed. The combination of spacer and multivalency effects led to a 38-fold increase in the potency of a divalent inhibitor with long spacer arms towards the PAO strain when compared with the free carbohydrate.

  1. Safety in the ARIES-III D-3He tokamak reactor design

    International Nuclear Information System (INIS)

    Herring, J.S.; Dolan, T.J.

    1992-01-01

    This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

  2. Turbine flow diagram of Paks-1 reactor

    International Nuclear Information System (INIS)

    Vancso, Tamas

    1983-01-01

    Computer calculations and programs are presented which inform the operators on the effect projected on the turbine and thermal efficiency of the modification in the flow diagram and in the starting parameters of the power cycle. In the program the expansion line of steam turbine type K-220-44 and the thermo-technical parameters of the elements of the feed-water heater system are determined. Detailed degree calculations for turbine unit of high pressure can be made. (author)

  3. The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

  4. Reversed field pinch reactor study 3

    International Nuclear Information System (INIS)

    Hollis, A.A.; Mitchell, J.T.D.

    1977-12-01

    This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

  5. Hungarian approach

    International Nuclear Information System (INIS)

    Hamar, K.

    1998-01-01

    This paper describes the licensing milestones of Paks NPP reactor protection refurbishment project starting from the simple task specification of high-tech I and C installation and up to acceptance tests and issuing license which are scheduled for 1999. Specific emphasis are put on the structure of the reactor protection refurbishment project licensing documentation

  6. Reliability analysis of protection systems in NPP applying fault-tree analysis method

    International Nuclear Information System (INIS)

    Bokor, J.; Gaspar, P.; Hetthessy, J.; Szabo, G.

    1998-01-01

    This paper demonstrates the applicability and limits of dependability analysis in nuclear power plants (NPPS) based on the reactor protection refurbishment project (RRP) in NPP Paks. This paper illustrates case studies from the reliability analysis for NPP Paks. It also investigates the solutions for the connection between the data acquisition and subsystem control units (TSs) and the voter units (VTs), it analyzes the influence of the voting in the VT computer level, it studies the effects of the testing procedures to the dependability parameters. (author)

  7. Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

    2003-01-01

    The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

  8. The World's Reactors no. 70 - Forsmark 3, BWR-75

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    A large pull-out wall chart is presented showing a coloured cut-away diagram of the Forsmark 3 station. It is accompanied by 2 small sketches one showing the layout of station buildings and the other the inside of the reactor vessel. Parameters are listed. (U.K.)

  9. New models in VERONA 7.0 system

    Energy Technology Data Exchange (ETDEWEB)

    Pos, Istvan; Kalya, Zoltan; Parko, Tamas [Paks Nuclear Power Plant Ltd, Paks (Hungary); Patai-Szabo, Sandor [TS Enercon Ltd., Budapest (Hungary)

    2016-09-15

    Nowadays the installation of a new modernized VERONA core monitoring system (version V7.0) is in process at the NPP Paks. The most important steps of the current improvements are as follows: complete replacement of the hardware and the local area network; application of a new operating system and ''virtual machine'' (VM) technology; implementation of a new human-system interface; and last but not least, introduction of improved reactor physics calculations. Basic novelty of the modernized core analysis is the application of general purpose graphical processing units (GPGPU) in the on-line core-follow module. This new technology has allowed of performing the real-time node-wise core analysis by standard Paks NPP core design codes HELIOS/C-PORCA. The present paper gives a brief overview of the system version (V7.0), focusing to the models of reactor physics and results of validation. Main characteristics of new approaches of the modified on-line reactor physics calculations are also described.

  10. The analysis of 14.8 percent cold leg break without the application of hydroaccumulators in the PMK-NHV test facility

    International Nuclear Information System (INIS)

    Szabados, L.; Ezsoel, Gy.; Perneczky, L.

    1990-12-01

    A series of reactor safety tests have been performed in the experimental reactor simulation facility PMK-NHV of the Paks Nuclear Power Plant, Hungary, with and without the use of hydroaccumulator (SIT) operation. 14.8 percent cold leg break simulation experiments are reported without SITs in action, and the measurement results were analyzed using the RELAP5/mod2 computer code. The description of the experiment is followed by the parameter variations and their analysis, together with an interpretation of the measurement results. The lessons from the LOCA simulation tests are evaluated. (R.P.) 10 refs.; 48 figs.; 3 tabs

  11. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

  12. Development of telerobotic systems for reactor decommissioning, (3)

    International Nuclear Information System (INIS)

    Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

    1991-01-01

    This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

  13. Estudo do centro de massa e estabilidade de quatro posturas básicas do Kung-fu Pak Hok

    OpenAIRE

    Miranda,Pedro Jeferson; Brinatti,André Maurício; Silva,Silvio Luiz Rutz da; Godoy,Marino Luiz Michelin

    2016-01-01

    Este trabalho trata da análise dos centros de massa e do cálculo da estabilidade das quatro posturas básicas do Kung Fu Pak Hok. Embora a biomecânica tenha surgido em 1960, a sua aplicação em artes marciais, como no Kung Fu ainda é pouco frequente. Apesar de haver estudos de movimentos do Kung Fu, não há trabalhos sobre o centro de massa e a estabilidade para as posturas mais básicas. Este trabalho como objetivo descrever o centro de massa e a estabilidade das quatro posturas mais básicas do ...

  14. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

    2007-11-01

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

  15. Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

    International Nuclear Information System (INIS)

    El-Sahlamy, N.M.

    2017-01-01

    One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

  16. The qualification of U3O8 as research reactor fuel

    International Nuclear Information System (INIS)

    Krull, W.

    1983-01-01

    This report summarizes the today knowledge of the qualification status of U 3 O 8 as low enriched ( 3 O 8 is so far qualified to start testing of ten (10) fuel elements with an U-density of 3.1 g U/cc in the FRG-2 research reactor. (orig.) [de

  17. Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

    International Nuclear Information System (INIS)

    Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

    2004-01-01

    The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

  18. Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Khan, Jahirul Haque

    2013-01-01

    The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

  19. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  20. 3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

    International Nuclear Information System (INIS)

    Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

    1974-01-01

    1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

  1. A robot-automated work site for repair of the Chinon A3 reactor

    International Nuclear Information System (INIS)

    Raynal, A.

    1987-01-01

    In 1982, following degradation due to corrosion of low-carbon steel by carbon dioxide gas, the utility undertook to repair some of the support structures at Chinon A3. This involved consolidation and reinforcing thermocouples and gas monitor pipeworks supports. A welding process was selected and the use of robots became indispensable because of the large number of components to be replaced (200 per outage). Two robots, supplied with tool heads and replacement components from outside the reactor were used. The robots and their servers were coordinated by a central computer and monitored by a closed circuit television system. Each repair operation was performed after ''training'' on a full-scale mockup of the top of the reactor reconstructed from telemetry of the real reactor dimensions. Since becoming operational in June 1986, the robots have accumulated over 20 000 hours of operation and seventy parts have been welded to the reactor. A 3D CAD system has been adapted to simulate the robots and analyse long trajectories in order to reduce robot learning time [fr

  2. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99). Elle est

  3. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

  4. Uranium-fuel thermal reactor benchmark testing of CENDL-3

    International Nuclear Information System (INIS)

    Liu Ping

    2001-01-01

    CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

  5. Investigations related to a one-piece removal of the reactor block in the frame of the JRR-3 reconstruction program

    International Nuclear Information System (INIS)

    Onishi, N.; Kanenari, A.; Futamura, Y.; Sakurai, H.; Suzuki, S.; Nagase, T.; Iwatani, A.; Otsubo, F.

    1987-01-01

    In the Japan Atomic Energy Research Institute (JAERI), an outdated research reactor (Japan Research Reactor No.3; JRR-3) was removed to a storage facility between October 14th and November 7th, 1986. The removal of the 2250-ton reactor block (10 x 10 x 10 m) was performed as a part of a program to replace the JRR-3's core (10-MW thermal) with an upgraded research reactor core. The heavy water and fuel elements were taken out from the JRR-3 before removal work began. The reactor block was raised about 3.7 meters, using a 12-cubic meter steel frame and a center-hole jack system. The reactor block was then transported horizontally about 34 meters on steel rails, using four 100-ton jacks, to a storage facility. Finally, the reactor block was lowered 14 meters into the storage facility. After the reactor block was stored, a new 20-MW thermal, light-water moderated and cooled JRR-3 core will be built, with criticality targeted for 1989

  6. Decommissioning of the BR3 pressurized-water reactor

    International Nuclear Information System (INIS)

    Massaut, V.

    1996-01-01

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

  7. Application of RELAP5-3D code for thermal analysis of the ADS reactor core

    International Nuclear Information System (INIS)

    Fernandes, Gustavo Henrique Nazareno

    2018-01-01

    Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

  8. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

  9. Safety Culture Evaluation at Research Reactors of Pakistan Atomic Energy Commission

    International Nuclear Information System (INIS)

    Qamar, M.A.; Saeed, A.; Shah, J.H.

    2016-01-01

    The concept of safety culture was presented by IAEA in document INSAG-4 (1991), delineated as “assembly of characteristics and attitudes in organizations and individuals which establish that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance”. The purpose of this paper is to describe the evaluation of safety culture at research reactors of the Pakistan Atomic Energy Commission (PAEC). Evaluating the safety culture of a particular organization poses some challenges which can be resolved by using safety culture evaluation models like those of Sachein (1992) and Harber-Barrier(1998). In PAEC, safety culture is the integral part of management system which not only promotes safety culture throughout the organization but also enhances its significance. To strengthen the safety culture, PAEC is also participating in a number of international and regional meetings of IAEA regarding safety culture. PAEC and the national regulator Pakistan Nuclear Regulatory Authority (PNRA) are also arranging workshops, peer reviews, sharing operational experiences and interacting with IAEA missions to enhance its capabilities in the field of safety culture. The Directorate General of Safety (DOS) is a corporate office of PAEC for safety and regulatory matters. DOS is in the process of implementing a program to evaluate safety culture at nuclear installations of PAEC to ensure that safety culture is included as a vital segment of the Integral Management System of the establishment. In this regard, training sessions and lectures on safety culture evaluation are normally conducted in PAEC for awareness and enhancement of the safety culture program. Safety culture is also addressed in PNRA Regulations like PAK-909 and PAK-913. In this paper we will focus on the safety culture evaluation in our research reactors, i.e., PARR-1 and PARR-2. The evaluation results will be based on observations, interviews of employees, group discussions

  10. Reduction of waste arising as an option for improvement of waste management systems at NPPs with WWER type reactors

    International Nuclear Information System (INIS)

    Dultchenko, A.; Mikolaitchouk, H.

    1995-01-01

    After the USSR breakdown Ukraine inherited five NPPs with 12 WWER type reactor units and 4 RBMK type reactor units and no selected disposal site for NPP operational waste and just a few waste treatment facilities which had not been licensed or certified and could not be considered as complying safety requirements and NPP needs. At the same time the lack of competent designer organizations in Ukraine and the overall economical situation including the payment crisis resulted in significant delays in the development of radioactive waste management infrastructure and brought to the foreground a reduction of waste arisings and implementation of waste recycling technologies. In order to evaluate efficiency of waste management systems at Ukrainian NPPs in comparison with current practices at western NPPs and fix main deficiencies and optimum upgrading measures the comparative analyses of waste management systems at Ukrainian NPPs was initiated within the R and D program supported by the Ukrainian State Committee for Nuclear and Radiation Safety (UkrSCNRS). In carrying out the analyses the results of IAEA Technical Assistance Regional project on Advice on Waste Management at WWER type Reactors were used. Taking into account an influence of the Chernobyl accident consequences on the waste management system of Chernobyl NPP the case of Chernobyl NPP was set apart and cannot be considered typical so the authors confine their analysis to the WWER type reactors. For the purposes of comparison the related information about Kozlodui, Paks, Loviisa and Russian NPPs provided under the above-mentioned IAEA Regional Project was used

  11. 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.)

  12. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    Edwards, Robert; Huang, Zhengyu

    2001-01-01

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

  13. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    International Nuclear Information System (INIS)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

    2015-01-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  14. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

    2015-07-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  15. Advances in Reactor physics, mathematics and computation. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

  16. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  17. Microbial rehabilitation of soils in the vicinity of former coking plants; Mikrobielle Sanierung von Kokereiboeden

    Energy Technology Data Exchange (ETDEWEB)

    Knackmuss, H J [Fraunhofer-Institut fuer Grenzflaechen- und Bioverfahrenstechnik, Stuttgart (Germany); Bryniok, D [Fraunhofer-Institut fuer Grenzflaechen- und Bioverfahrenstechnik, Stuttgart (Germany)

    1997-12-31

    Two airlift reactors with a nominal volume of 15 liters were provided with a closed aeration circuit. This mode of operation for the first time permitted to determine the carbon balance of PAH degradation. A mineralisation rate of approx. 35% was found by this method, whereas in experiments performed in shaking bottles mineralisation was always over 60% in the case of PAH mixtures. Use of PAH mixtures leads to competitive effects. These effects were studied by means of bacterial pure cultures. Further fundamental studies were performed to find suitable solvents for PAH degradation in a culture system with two liquid phases and examine liquid-liquid extraction of PAH from soil washing water. (orig./SR) [Deutsch] Zwei Airliftreaktoren mit einem Nennvolumen von 15 Litern wurden mit einem geschlossenen Belueftungskreislauf versehen. Diese Betriebsweise erlaubte erstmals die Bestimmung einer Kohlenstoffbilanz des PAK-Abbaus. Diese ergab eine Mineralisation von ca. 35%, waehrend die Mineralisationsrate bei Versuchen im Schuettelkolben selbst im Falle von PAK-Gemischen immer ueber 60% lag. Bei der Verwertung von PAK-Gemischen treten Kompetitionseffekte auf. Diese wurden mit bakteriellen Reinkulturen untersucht. Weitere grundlegende Arbeiten betrafen die Auswahl geeigneter Loesungsmittel fuer den PAK-Abbau in einem Kultursystem mit zwei Fluessigphasen und die Fluessig/Fluessig-Extraktion von PAK aus Bodenwaschwasser. (orig./SR)

  18. Microbial rehabilitation of soils in the vicinity of former coking plants; Mikrobielle Sanierung von Kokereiboeden

    Energy Technology Data Exchange (ETDEWEB)

    Knackmuss, H.J. [Fraunhofer-Institut fuer Grenzflaechen- und Bioverfahrenstechnik, Stuttgart (Germany); Bryniok, D. [Fraunhofer-Institut fuer Grenzflaechen- und Bioverfahrenstechnik, Stuttgart (Germany)

    1996-12-31

    Two airlift reactors with a nominal volume of 15 liters were provided with a closed aeration circuit. This mode of operation for the first time permitted to determine the carbon balance of PAH degradation. A mineralisation rate of approx. 35% was found by this method, whereas in experiments performed in shaking bottles mineralisation was always over 60% in the case of PAH mixtures. Use of PAH mixtures leads to competitive effects. These effects were studied by means of bacterial pure cultures. Further fundamental studies were performed to find suitable solvents for PAH degradation in a culture system with two liquid phases and examine liquid-liquid extraction of PAH from soil washing water. (orig./SR) [Deutsch] Zwei Airliftreaktoren mit einem Nennvolumen von 15 Litern wurden mit einem geschlossenen Belueftungskreislauf versehen. Diese Betriebsweise erlaubte erstmals die Bestimmung einer Kohlenstoffbilanz des PAK-Abbaus. Diese ergab eine Mineralisation von ca. 35%, waehrend die Mineralisationsrate bei Versuchen im Schuettelkolben selbst im Falle von PAK-Gemischen immer ueber 60% lag. Bei der Verwertung von PAK-Gemischen treten Kompetitionseffekte auf. Diese wurden mit bakteriellen Reinkulturen untersucht. Weitere grundlegende Arbeiten betrafen die Auswahl geeigneter Loesungsmittel fuer den PAK-Abbau in einem Kultursystem mit zwei Fluessigphasen und die Fluessig/Fluessig-Extraktion von PAK aus Bodenwaschwasser. (orig./SR)

  19. Preparations for the shipment of RA-3 reactor irradiated fuel

    International Nuclear Information System (INIS)

    Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

    2002-01-01

    During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

  20. Reading the Af-Pak Narrative, from the US Disengagement to Russian Re-Engagement

    Directory of Open Access Journals (Sweden)

    A. Dhaka

    2017-01-01

    Full Text Available The US has prolonged its stay in Afghanistan with the security situation remaining far from improving. The indefatigable demand for resources to maintain counter-insurgency operations was a major debate in 2016 US Presidential elections with a demand for an earlier withdrawal from America’s trillion dollars plus war effort. Russians having sensed the weakening of the US infl uence warmed upto the idea of new Afghan situation involving Taliban and their masters, the Pakistan army. Russia had experienced vulnerabilities of Islamisation in Central Asia and Caucasus, and the ISIS brand radicalisation added to the fear of political destabilisation of Central Asian states. The Islamic State showed up in Afghanistan and Pakistan as ISIS-Khorasan branch. Russia needed Pakistan as an ally to fi ght Daesh’s presence on its southern periphery. However, there remained many intertwined security challenges that complicate the South Asian geopolitics, especially, the Af-Pak region. Russia’s Taliban policy might be the hitherto unused leverage that it might be using in order to strike balance all along the shatter belt.

  1. Återvinning av dryckeskartonger : En studie som syftar till att öka återvinningsgraden av Tetra Paks förpackningar i Indonesien

    OpenAIRE

    Backlund, Per

    2014-01-01

    People of the modern world consume more than they ever used to do. Because of the close correlation between consumption and the amount of waste, the waste volume is also expected to increase. The purpose of this study is to examine if some measures in the recycling process from Sweden could be implemented in Indonesia. In fact, Indonesia is one of the countries in which the waste management system is struggling. Tetra Pak, one of the world leading producer of food packaging, is studied in thi...

  2. Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.

    1994-01-01

    TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

  3. On the major DYN3D developments for fast reactor design and transient analysis

    International Nuclear Information System (INIS)

    Merk, B.; Kliem, S.

    2013-01-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  4. On the major DYN3D developments for fast reactor design and transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

    2013-07-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  5. 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.

  6. RA reactor exploitation, task 3.08/01; Zadatak 3.08/01 - Eksploatacija reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report.

  7. Operating reactors licensing actions summary. Vol. 3, No. 6

    International Nuclear Information System (INIS)

    1983-07-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  8. Study of seismic responses of Candu-3 reactor building using isolator bearings

    International Nuclear Information System (INIS)

    Biswas, J.K.

    1992-01-01

    Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

  9. 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

  10. Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

    Science.gov (United States)

    Longoni, Gianluca; Anderson, Stanwood L.

    2009-08-01

    The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

  11. Research on intelligent monitor for 3D power distribution of reactor core

    International Nuclear Information System (INIS)

    Xia, Hong; Li, Bin; Liu, Jianxin

    2014-01-01

    Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

  12. Coastal erosion and accretion in Pak Phanang, Thailand by GIS analysis of maps and satellite imagery

    Directory of Open Access Journals (Sweden)

    Sayedur Rahman Chowdhury

    2013-12-01

    Full Text Available Coastal erosion and accretion in Pak Phanang of southern Thailand between 1973 and 2003 was measured using multi-temporal topographic maps and Landsat satellite imageries. Within a GIS environment landward and seaward movements of shoreline was estimated by a transect-based analysis, and amounts of land accretion and erosion were estimated by a parcel-based geoprocessing. The whole longitudinal extent of the 58 kilometer coast was classified based on the erosion and accretion trends during this period using agglomerative hierarchical clustering approach. Erosion and accretion were found variable over time and space, and periodic reversal of status was also noticed in many places. Estimates of erosion were evaluated against field-survey based data, and found reasonably accurate where the rates were relatively great. Smoothing of shoreline datasets was found desirable as its impacts on the estimates remained within tolerable limits.

  13. Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

    2017-01-01

    Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

  14. G2 and G3 reactors design; Description des reacteurs G2 et G3

    Energy Technology Data Exchange (ETDEWEB)

    Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

  15. 3. International conference on catalysis in membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

  16. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    International Nuclear Information System (INIS)

    Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

    2008-01-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  17. Elevated p21-Activated Kinase 2 Activity Results in Anchorage-Independent Growth and Resistance to Anticancer Drug–Induced Cell Death

    Directory of Open Access Journals (Sweden)

    Jerry W. Marlin

    2009-03-01

    Full Text Available p21-Activated kinase 2 (PAK-2 seems to be a regulatory switch between cell survival and cell death signaling. We have shown previously that activation of full-length PAK-2 by Rac or Cdc42 stimulates cell survival, whereas caspase activation of PAK-2 to the proapoptotic PAK-2p34 fragment is involved in the cell death response. In this study, we present a role of elevated activity of full-length PAK-2 in anchorage-independent growth and resistance to anticancer drug–induced apoptosis of cancer cells. Hs578T human breast cancer cells that have low levels of PAK-2 activity were more sensitive to anticancer drug–induced apoptosis and showed higher levels of caspase activation of PAK-2 than MDA-MB435 and MCF-7 human breast cancer cells that have high levels of PAK-2 activity. To examine the role of elevated PAK-2 activity in breast cancer, we have introduced a conditionally active PAK-2 into Hs578T human breast cells. Conditional activation of PAK-2 causes loss of contact inhibition and anchorage-independent growth of Hs578T cells. Furthermore, conditional activation of PAK-2 suppresses activation of caspase 3, caspase activation of PAK-2, and apoptosis of Hs578T cells in response to the anticancer drug cisplatin. Our data suggest a novel mechanism by which full-length PAK-2 activity controls the apoptotic response by regulating levels of activated caspase 3 and thereby its own cleavage to the proapoptotic PAK-2p34 fragment. As a result, elevated PAK-2 activity interrupts the apoptotic response and thereby causes anchorage-independent survival and growth and resistance to anticancer drug–induced apoptosis.

  18. Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

    International Nuclear Information System (INIS)

    Longoni, G.

    2010-01-01

    The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBand R interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

  19. The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

    International Nuclear Information System (INIS)

    Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

    2003-01-01

    Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

  20. Technical report on implementation of reactor internal 3D modeling and visual database system

    International Nuclear Information System (INIS)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun

    1996-06-01

    In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation's integrated computer aided engineering system, such as Mitsubishi's NUWINGS (Japan), AECL's CANDID (Canada) and Duke Power's PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new

  1. Technical report on implementation of reactor internal 3D modeling and visual database system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

  2. 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

  3. Seismic safety programme at NPP Paks. Propositions for coordinated international activity in seismic safety of the WWER-440 V-213

    International Nuclear Information System (INIS)

    Katona, T.

    1995-01-01

    This paper presents the Paks NPP seismic safety program, highlighting the specifics of the WWER-440/213 type in operation, and the results of work obtained so far. It covers the following scope: establishment of the seismic safety program (original seismic design, current requirements, principles and structure of the seismic safety program); implementation of the seismic safety program (assessing the seismic hazard of the site, development of the new concept of seismic safety for the NPP, assessing the seismic resistance of the building and the technology); realization of the seismic safety of higher level (technical solutions, drawings, realization); ideas and propositions for coordinated international activity

  4. Separation properties of aluminium-plastic laminates in post-consumer Tetra Pak with mixed organic solvent.

    Science.gov (United States)

    Zhang, S F; Zhang, L L; Luo, K; Sun, Z X; Mei, X X

    2014-04-01

    The separation properties of the aluminium-plastic laminates in postconsumer Tetra Pak structure were studied in this present work. The organic solvent blend of benzene-ethyl alcohol-water was used as the separation reagent. Then triangle coordinate figure analysis was taken to optimize the volume proportion of various components in the separating agent and separation process. And the separation temperature of aluminium-plastic laminates was determined by the separation time, efficiency, and total mass loss of products. The results show that cost-efficient separations perform best with low usage of solvents at certain temperatures, for certain times, and within a certain range of volume proportions of the three components in the solvent agent. It is also found that similar solubility parameters of solvents and polyethylene adhesives (range 26.06-34.85) are a key factor for the separation of the aluminium-plastic laminates. Such multisolvent processes based on the combined-system concept will be vital to applications in the recycling industry.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-04-01

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

  6. Development of a 3-D flow analysis computer program for integral reactor

    International Nuclear Information System (INIS)

    Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

    2003-01-01

    A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

  7. 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

  8. Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

    International Nuclear Information System (INIS)

    Chi-Mei Lee; Kuo-Tsang Hung

    2006-01-01

    The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

  9. Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

    International Nuclear Information System (INIS)

    1992-04-01

    Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

  10. A review of cancer mortality data of radiation workers of Nuclear Power Plant, Paks, Hungary, in the light the international radiation epidemiology study

    International Nuclear Information System (INIS)

    Turai, I.; Kerekes, A.; Otos, M.; Veress, K.

    2007-01-01

    Complete text of publication follows. Objective: To give a review of cancer mortality data among Hungarian radiation workers in nuclear industry in comparison with the results of the international nuclear workers' study prevailing the size of the study group of all former studies. Methods: Retrospective cohort study including 598,068 workers of 154 nuclear establishments in 15 countries (AUS, BEL, CAN, FIN, FRA, GER, HUN, JAP, LIT, ROK, SLK, SPA, SWE, UK, USA) coordinated by the International Agency for Research on Cancer (IARC, Lyon, France). The national study was extended for an additional 4-year period. Results: In the international study 407,391 persons in 13 years of average employment received 19.4 mSv mean cumulative dose, while in the national study 3322 radiation workers of Nuclear Power Plant (NPP) Paks, Hungary, in 14 years of follow-up period accumulated in average 5.13 mSv, only. There were 5233 cancer deaths registered in the international study, associated with an estimated ERR of 0.97 per Sv. Thus, 19.4 mSv recorded cumulative dose can explain 1 to 2% of cancer death cases. In radiation workers of NPP, Paks, during the period of 1985-1998 there were 40 cancer deaths observed against the expected 58.8 cases. In a further four year period (1999-2002) 29 cancer death cases were identified vs. the expected 65.5 cases. The SMR for the cancer death cases registered in recent and former radiation workers of NPP, Paks in the 18-year follow-up period is 56%. The SMR from all causes was even lower, 40% only. Conclusions: In the international study the mean accumulated radiation dose received by nuclear workers in 13 years is below of the recent annual dose limit (20 mSv/yr of the effective dose). The average value for the whole of radiation workers in 15 countries is almost 4-times higher of that registered in Hungary. The 'healthy worker effect' in the nuclear industry, and particularly in Hungary has been proven, once again. Nevertheless, the results

  11. The ARIES-III D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1992-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

  12. Annealing of the BR3 reactor pressure vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

    1985-01-01

    The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

  13. 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

  14. Croatian-Hungarian cooperation on the Danube river radioactivity measurements

    International Nuclear Information System (INIS)

    Lulic, S.; Vancsura, P.

    2003-01-01

    Danube river radioactivity measurements on the border profile Mohac-Batina have been performed since the beginning of 1978 with varying frequency of sampling. Thus, in the period before nuclear power plant Paks started to work joint croatian-hungarian sampling at the border profile was taking place four times a year; the obtained results of measured radioactivity levels were used to assess radioactivity background data. From the start of nuclear power plant Paks running until Chernobyl reactor accident (April 1986) sampling was performed six times a year. After the Chernobyl accident, samples have been taken every month. Since decreased Chernobyl reactor accident influence was estimated until present samples have been taken six times a year. On the Danube river border profile the concentration activity of gamma radionuclides has been determined in water samples (filtered water and suspended matter), and in fish, sediment and Danube river algae samples. (authors)

  15. Reliability analysis of the automatic control and power supply of reactor equipment

    International Nuclear Information System (INIS)

    Monori, Pal; Nagy, J.A.; Meszaros, Zoltan; Konkoly, Laszlo; Szabo, Antal; Nagy, Laszlo

    1988-01-01

    Based on reliability analysis the shortcomings of nuclear facilities are discovered. Fault tree types constructed for the technology of automatic control and for power supply serve as input data of the ORCHARD 2 computer code. In order to charaterize the reliability of the system, availability, failure rates and time intervals between failures are calculated. The results of the reliability analysis of the feedwater system of the Paks Nuclear Power Plant showed that the system consisted of elements of similar reliabilities. (V.N.) 8 figs.; 3 tabs

  16. 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)

  17. 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

  18. 3 D flow computations under a reactor vessel closure head

    International Nuclear Information System (INIS)

    Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

    1995-12-01

    The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

  19. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  20. Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

    2012-03-15

    JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

  1. Summary of research performed since June 1996 and final summary

    International Nuclear Information System (INIS)

    Masopust, R.

    1999-01-01

    Significant influence of large relative displacement might be caused by faulting-induced movements, lateral spreading (not expected in Paks), liquefaction induced ground movements (not expected in Paks), landslides and slope failures (not expected in Paks), settlement and seismic-induced motions and deformations of the building structure into the which the buried pipe is attached (compensated by flexible connections in Paks). Dynamic amplification does not play an important role in the response of buried pipes. Only the static response of buried pipelines when subjected to propagate seismic waves is important when the large relative displacements of the ground along the pipeline cannot occur. This presentation covers seismic evaluation of buried pipes with emphasis on the main emergency water supply system and the seismic margin assessment applied to large vertical cylindrical flat bottom tanks installed inside the main building of NPP. Conclusions derived from the obtained results are as follows: no seismic upgrading necessary for vertical single and also multi cylindrical tanks located inside the main reactor building mostly on the elevation ± 0.00 m; sliding shear capacity of such tanks when they stand on special grids without any anchorage needs more detailed investigation

  2. Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

    International Nuclear Information System (INIS)

    Koch, R.; Findeisen, A.

    1986-04-01

    The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

  3. Task 7a - dynamic analysis of Paks NPP structures: Reactor building

    Energy Technology Data Exchange (ETDEWEB)

    Zola, M [ISMES S.p.A., Bergamo (Italy)

    1995-07-01

    This report refers to the activities of a sub-contract to the Project RER/9/046, awarded to ISMES by the International Atomic Energy Agency (IAEA) of Vienna, to compare the results obtained from the experimental activities performed under previous contract by ISMES with those coming from analytical studies performed in the framework of the Coordinated Research Programme (CRP) on 'Benchmark Study for the Seismic Analysis and Testing of WWER-type Nuclear Power Plants' by other Institutions. After a brief introduction to the problem in Chapter 1, the identification of the comparison positions and reference directions is given in Chapter 3. A very quick description of the performed experimental tests is given in Chapter 4, whereas the characteristics of both experimental and analytical data are presented in Chapter 5. The data processing procedures are reported in Chapter 6 and some simple remarks are given in Chapter 7. (author)

  4. Task 7a - dynamic analysis of Paks NPP structures: Reactor building

    International Nuclear Information System (INIS)

    Zola, M.

    1995-01-01

    This report refers to the activities of a sub-contract to the Project RER/9/046, awarded to ISMES by the International Atomic Energy Agency (IAEA) of Vienna, to compare the results obtained from the experimental activities performed under previous contract by ISMES with those coming from analytical studies performed in the framework of the Coordinated Research Programme (CRP) on 'Benchmark Study for the Seismic Analysis and Testing of WWER-type Nuclear Power Plants' by other Institutions. After a brief introduction to the problem in Chapter 1, the identification of the comparison positions and reference directions is given in Chapter 3. A very quick description of the performed experimental tests is given in Chapter 4, whereas the characteristics of both experimental and analytical data are presented in Chapter 5. The data processing procedures are reported in Chapter 6 and some simple remarks are given in Chapter 7. (author)

  5. Removal of the damaged fuel from Paks-2 pit

    International Nuclear Information System (INIS)

    Cserhati, A.

    2007-01-01

    On 10 April 2003, during the outage period a chemical cleaning program for the fuel assemblies has been carried out at the unit 2, in a specially designed cleaning tank. The tank is located in a pit, near to the reactor. 30 fuel assemblies have been significantly damaged due to inadequate cooling. After the extensive preparation - lasting 3,5 years - the pickup and encapsulation of the damaged fuel has been preformed. All tasks have been carried out safely, during the planned 3 months without any substantial problems. This paper covers the events of this last implementation phase. The main topics are: initial conditions of the pit and the cleaning tank before the start of the recovery; tasks and responsibilities, organization, timing, control.; visual following for the fuel removal; technology features, steps made; short and long term tasks after the removal of the fuel; summary, achievements. (author)

  6. Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

    International Nuclear Information System (INIS)

    Scari, Maria Elizabeth

    2017-01-01

    Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

  7. 3D CFD for chemical transport profiles in a rotating disk CVD reactor

    Science.gov (United States)

    Han, Jong-Hyun; Yoon, Do-Young

    2010-06-01

    The RDCVD (Rotating Disk Chemical Vapor Deposition) technique is an appropriate method for uniform deposition of grains, such as compound semiconductior materials. The substrate temperature and rotation speed are the major factors, which determine the thickness uniformity of the deposited films. This paper investigates 3D CFD (3 Dimensional Computational Fluid Dynamics) simulation results of flow and heat transfer in a reactor of RDCVD using Fluent. In order to establish the reducibility of buoyancy effect on deposition quality, the chemical transport profile upon the disk heated is examined successfully in 3D domain for different rotating speeds. The resulting vortex flows due the simultaneous buoyance and centrifuge are discussed qualitatively in the 3D virtual system of a RDCVD reactor. 3D CFD is even more effective to describe the internal vortex flows due to the competitive inlet, buoyancy and centrifuge flows, which cannot be realized in the general 2D (2 Dimensional) CFD.[Figure not available: see fulltext.

  8. Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

    2012-07-15

    Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

  9. Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

    KAUST Repository

    Karsenty, Florent

    2012-08-16

    Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

  10. Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

    International Nuclear Information System (INIS)

    Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

    2015-01-01

    Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

  11. Study on effects of development of reactor constant in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-12-01

    Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

  12. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  13. Main features of the pressurized component life management related R and D activity in Hungary

    International Nuclear Information System (INIS)

    Gillemot, F.; Oszwald, F.

    1995-01-01

    During the last one and half years the following main developments related to the life-time management of the NPP units took place in Hungary: 1. National project named AGNES (Advanced General New Evaluation of Safety) finished. 2. The first results of the extended surveillance program of NPP Paks has been measured. 3. The upgraded ultrasonic testing equipment used for RPV outside testing, and nozzle testing is regularly used in refuelling periods. 4. Radiation embrittlement and thermal ageing research started on RPV materials, mainly on cladding. 5. Participation in the development of the IAEA RPV ageing database. 6. Participation in IAEA pilot studies on ageing. 7. Operation of the Budapest Research Reactor, and building a new high capacity helium cooled irradiation rig. 8. Nondestructive evaluation of material ageing of a steam generator vessels. 9. A life-time calculation of Paks RPV-s. This report is a short survey of these developments. 19 refs, 4 figs

  14. Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

    International Nuclear Information System (INIS)

    2015-01-01

    This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

  15. Extension of the GeN-Foam neutronic solver to SP3 analysis and application to the CROCUS experimental reactor

    International Nuclear Information System (INIS)

    Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

    2017-01-01

    Highlights: • Development and verification of an SP 3 solver based on OpenFOAM. • Integration into the GeN-Foam multi-physics platform. • Application of the new GeN-Foam SP 3 solver to the CROCUS reactor. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and at the EPFL has been developing since 2013 a multi-physics platform for coupled reactor analysis named GeN-Foam. The developed tool includes a solver for the eigenvalue and transient solution of multi-group neutron diffusion equations. Although frequently used in reactor analysis, the diffusion theory shows some limitations for core configurations involving strong anisotropies, which is the case for the CROCUS research reactor at the EPFL. The use of an SP 3 approximation to neutron transport can often lead to visible improvements in a code predictive capabilities, especially for one-directional anisotropies, with acceptable added computational cost vs diffusion. Following some modelling issues for the CROCUS reactor, and in order to improve the GeN-Foam modelling capabilities, the GeN-Foam diffusion solver has been extended to allow for SP 3 analyses. The present paper describes such extension and a preliminary verification using a mini-core PWR benchmark. The newly developed solver is then applied to the analysis of the CROCUS experimental reactor and results are compared to Monte Carlo calculations, as well as to the results of the diffusion solver.

  16. 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.

  17. 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.

  18. 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

  19. Tritium production, management and its impact on safety for a D-3He fusion reactor

    International Nuclear Information System (INIS)

    Sze, D.K.; Herring, S.; Sawan, M.

    1991-11-01

    About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

  20. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    International Nuclear Information System (INIS)

    Loika, E.F.

    1994-01-01

    This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

  1. 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.

  2. 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

  3. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  4. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1998-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  5. Regulatory Approach To and Lessons Learned with Licensing of Service Life Extension at PAKS NPP

    International Nuclear Information System (INIS)

    Petofi, G.

    2012-01-01

    Paks Nuclear Power Plant of Hungary decided to extend the original design lifetime of the plant by 20 years, which expires on December 31, 2012 concerning unit 1. The Hungarian Atomic Energy Authority established the legal environments in order to license the extension using an approach similar to that followed in the United States. The regulation specifies the pre-conditions for the extension, defines the scoping and screening process for the passive and long lived systems, structures and components to be involved in the licensing and the respective methods of treatment and also determines how the active components shall be dealt with during the extended lifetime. The regulatory procedure is a two-step process including the oversight of the preparatory programme of the operator for the extension that started 4 years before the expiry of the lifetime and the licensing process itself, which is currently under way for unit 1 after the submittal of the licensee's application at the end of 2011. The first experiences with the regulatory assessment of the application are available yet and presented in this paper. (author)

  6. Characteristics of UV-MicroO3 Reactor and Its Application to Microcystins Degradation during Surface Water Treatment

    Directory of Open Access Journals (Sweden)

    Guangcan Zhu

    2015-01-01

    Full Text Available The UV-ozone (UV-O3 process is not widely applied in wastewater and potable water treatment partly for the relatively high cost since complicated UV radiation and ozone generating systems are utilized. The UV-microozone (UV-microO3, a new advanced process that can solve the abovementioned problems, was introduced in this study. The effects of air flux, air pressure, and air humidity on generation and concentration of O3 in UV-microO3 reactor were investigated. The utilization of this UV-microO3 reactor in microcystins (MCs degradation was also carried out. Experimental results indicated that the optimum air flux in the reactor equipped with 37 mm diameter quartz tube was determined to be 18∼25 L/h for efficient O3 generation. The air pressure and humidity in UV-microO3 reactor should be low enough in order to get optimum O3 output. Moreover, microcystin-RR, YR, and LR (MC-RR, MC-YR, and MC-LR could be degraded effectively by UV-microO3 process. The degradation of different MCs was characterized by first-order reaction kinetics. The pseudofirst-order kinetic constants for MC-RR, MC-YR, and MC-LR degradation were 0.0093, 0.0215, and 0.0286 min−1, respectively. Glucose had no influence on MC degradation through UV-microO3. The UV-microO3 process is hence recommended as a suitable advanced treatment method for dissolved MCs degradation.

  7. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  8. Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

    International Nuclear Information System (INIS)

    Behar, Christophe

    2014-01-01

    Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

  9. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  10. Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

    International Nuclear Information System (INIS)

    Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

    2008-01-01

    The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

  11. IVO participation in IAEA benchmark for VVER-type nuclear power plants seismic analysis and testing

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1997-12-01

    This study is a part of the IAEA coordinated research program 'Benchmark study for the Seismic Analysis and Testing of VVER Type NPPs'. The study reports the numerical simulation of the blast test for Paks and Kozloduy nuclear power plants beginning from the recorded free-field response and computing the structural response at various points inside the reactor building. The full-scale blast tests of the Paks and Kozloduy NPPs took place in December 1994 and in July 1996. During the tests the plants operated normally. The instrumentation for the tests consisted of 52 recording channels with 200 Hz sampling rate. Detonating 100 kg charges in 50-meter deep boreholes at 2.5-km distance from the plant carried out the blast tests. The 3D structural models for both reactor buildings were analyzed in the frequency domain. The number of modes extracted in both cases was about 500 and the cut-off frequency was 25 Hz. In the response history run the responses of the selected points were evaluated. The input values for response history run were the three components of the excitation, which were transformed from time domain to the frequency domain with the aid of Fourier transform. The analysis was carried out in frequency domain and responses were transferred back to time domain with inverse Fourier transform. The Paks and Kozloduy blast tests produced a wealth of information on the behavior of the nuclear power plant structures excited by blast type loads containing also the low frequency wave train if albeit with small energy content. The comparison of measured and calculated results gave information about the suitability of the selected analysis approach for the investigated blast type loading

  12. Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

    Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

  13. Autism-like Deficits in Shank3-Deficient Mice Are Rescued by Targeting Actin Regulators

    Directory of Open Access Journals (Sweden)

    Lara J. Duffney

    2015-06-01

    Full Text Available Haploinsufficiency of the Shank3 gene, which encodes a scaffolding protein at glutamatergic synapses, is a highly prevalent and penetrant risk factor for autism. Using combined behavioral, electrophysiological, biochemical, imaging, and molecular approaches, we find that Shank3-deficient mice exhibit autism-like social deficits and repetitive behaviors, as well as the significantly diminished NMDA receptor (NMDAR synaptic function and synaptic distribution in prefrontal cortex. Concomitantly, Shank3-deficient mice have a marked loss of cortical actin filaments, which is associated with the reduced Rac1/PAK activity and increased activity of cofilin, the major actin depolymerizing factor. The social deficits and NMDAR hypofunction are rescued by inhibiting cofilin or activating Rac1 in Shank3-deficient mice and are induced by inhibiting PAK or Rac1 in wild-type mice. These results indicate that the aberrant regulation of synaptic actin filaments and loss of synaptic NMDARs contribute to the manifestation of autism-like phenotypes. Thus, targeting actin regulators provides a strategy for autism treatment.

  14. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

  15. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

  16. Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

    International Nuclear Information System (INIS)

    González Mantecón, Javier

    2015-01-01

    The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

  17. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

    International Nuclear Information System (INIS)

    2014-02-01

    JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

  18. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

    International Nuclear Information System (INIS)

    2007-03-01

    In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

  19. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

    International Nuclear Information System (INIS)

    2009-01-01

    In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

  20. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-02-15

    JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

  1. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

    2013-01-01

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  2. Development of a new WWER-440 fuel design

    International Nuclear Information System (INIS)

    Coucil, D.; Totev, T.

    1998-01-01

    In March 1996 British Nuclear Fuel Limited signed a contract with Imatran Voima and Paks Nuclear Power Plant to design, develop, license and supply 5 Lead Test Assemblies to the WWER-440 reactor at Loviisa in Finland. In June 1998 the manufacture of these 5 assemblies (4 fixed assemblies and 1 follower assembly) was completed. The fuel is expected to be loaded into Loviisa Unit 2 reactor during the shutdown scheduled for September of this year. (Authors)

  3. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  4. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

  5. Plutonium-burn high temperature gas-cooled reactor for 3E+3S

    International Nuclear Information System (INIS)

    Okamoto, Koji

    2015-01-01

    The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

  6. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst

    Directory of Open Access Journals (Sweden)

    New Pei Yee

    2008-04-01

    Full Text Available A one-dimensional mathematical model was developed to simulate the performance of catalytic fixed bedreactor for carbon dioxide reforming of methane over Rh/Al2O3 catalyst at atmospheric pressure. The reactionsinvolved in the system are carbon dioxide reforming of methane (CORM and reverse water gas shiftreaction (RWGS. The profiles of CH4 and CO2 conversions, CO and H2 yields, molar flow rate and molefraction of all species as well as reactor temperature along the axial bed of catalyst were simulated. In addition,the effects of different reactor temperature on the reactor performance were also studied. The modelscan also be applied to analyze the performances of lab-scale micro reactor as well as pilot-plant scale reactorwith certain modifications and model verification with experimental data. © 2008 BCREC UNDIP. All rights reserved.[Received: 20 August 2008; Accepted: 25 September 2008][How to Cite: N.A.S. Amin, I. Istadi, N.P. Yee. (2008. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst. Bulletin of Chemical Reaction Engineering and Catalysis, 3 (1-3: 21-29. doi:10.9767/bcrec.3.1-3.19.21-29

  7. An assesment of the characteristics of the GM detectors and iodine remote detectors of the Paks environmental monitoring system based on the data measured from 1982 to 1985

    International Nuclear Information System (INIS)

    Nagy, Gy.; Lang, Edit; Deme, S.; Feher, I.

    1986-03-01

    Measurements performed at the GM detectors and iodine remote detectors of the continuous environmental monitoring system of the Paks NPP can be used for estimating the effect of atmospheric releases. Based on the investigations carried out from Sep. 1982 to July 1985, a good correlation between the signals and the background radioactivity levels could be established. It was further stated that radon fallout during raining was responsible for significant signal changes of both types of detectors. (V.N.)

  8. Monitoring and assessment of health issues at energy plant and gas station Pak steel bin Qasim Karachi

    International Nuclear Information System (INIS)

    Memon, Z.

    2005-01-01

    No doubt Environmental and health safety issues in big cities of Pakistan are developing havoc problems due to mechanized operations by emitting flue gases, effluent and acoustic noise, which it is my topic to discuss in detail. Acoustic noise is one of the major environmental problems in Industrial Plants. The noise study under taken in detail at feed pumps, super heater, exhausters and accumulators of Energy plant (E.P) as the regulators, control room etc. of Gas station (G.S) Pak Steel Bin Qasim Karachi. In light of permissible occupational noise exposure limits, as allowed by the ISO,EEC and other National Standards, some recommendations have been made to provide safety measures for workers against high level noise health hazards like head ache, hearing problem, Irritation, accidents at work, tension, disturbance to work and so many physiological and psychological effects, along with guidelines to overcome the break downs an improve efficiency of the plants. (Orig./A.B.)

  9. Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

    International Nuclear Information System (INIS)

    Guidez, J.; Saturnin, A.

    2016-01-01

    In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

  10. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems

  11. Verification of the enrichment of fresh VVER-440 fuel assemblies at NPP Paks

    Energy Technology Data Exchange (ETDEWEB)

    Almasia, I.; Hlavathya, Z.; Nguyena, C. T. [Institute of Isotopes, Hungarian Academy of Sciences, Budapest, (Hungary); others, and

    2012-06-15

    A Non Destructive Analysis (NDA) method was developed for the verification of {sup 235}U enrichment of both homogeneous and profiled VVER-440 reactor fresh fuel assemblies by means of gamma spectrometry. A total of ca. 30 assemblies were tested, five of which were homogeneous, with {sup 235}U enrichment in the range 1,6% to 3,6%, while the others were profiled with pins of 3,3% to 4,4% enrichment. Two types of gamma detectors were used for the test measurements: 2 coaxial HPGe detectors and a miniature CdZnTe (CZT) detector fitting into the central tube of the assemblies. It was therefore possible to obtain information from both the inside and the outside of the assemblies. It was shown that it is possible to distinguish between different types of assemblies within a reasonable measurement time (about 1000 sec). For the HPGe measurements the assemblies had to be lifted out from their storage rack, while for the CZT detector measurements the assemblies could be left at their storage position, as it was shown that the neighbouring assemblies do not affect measurement inside the assemblies' central tube. The measured values were compared to Monte Carlo simulations carried out using the MCNP code, and a recommendation for the optimal approach to verify the {sup 235}U enrichment of fresh VVER-440 reactor fuel assemblies is suggested.

  12. 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

  13. 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)

  14. Reactor safety study applied to the Forsmark 3 Power Plant

    International Nuclear Information System (INIS)

    Ericsson, G.; Tiren, L.I.

    1978-01-01

    A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

  15. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2009-01-01

    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

  16. Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E

    1985-07-01

    A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

  17. Conditions of job performance quality during nuclear reactor maintenance

    International Nuclear Information System (INIS)

    Babos, K.; Takacs, J.

    1991-01-01

    The education and training programs of the maintenance personnel at the Paks Nuclear Power Plant, Hungary are described. In addition to well trained personnel, the conditions of job quality involves the preparation of job by proper check lists and availability of documentation, and the inspection and checking of the accomplishment of job tasks. Such an inspection procedure is proposed. (R.P.) 2 figs

  18. The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

    The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

  19. Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

    International Nuclear Information System (INIS)

    Kishimoto, Katsumi; Arigane, Kenji

    2005-03-01

    Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18 (case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes. Total activation activity of reactor by the revaluation depended on control rods, thermal shield plates and horizontal experimental tubes, and the value after 1 year from the permanent shutdown of reactor was 1.9x10 14 Bq. (author)

  20. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  1. Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

    International Nuclear Information System (INIS)

    Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

    2005-01-01

    Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

  2. Revision of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    Takano, Hideki; Kaneko, Kunio.

    1989-10-01

    To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

  3. Caspase Activation of p21-Activated Kinase 2 Occurs During Cisplatin-Induced Apoptosis of SH-SY5Y Neuroblastoma Cells and in SH-SY5Y Cell Culture Models of Alzheimer’s and Parkinson’s Disease

    Directory of Open Access Journals (Sweden)

    Jerry W. Marlin

    2010-04-01

    Full Text Available p21-activated kinase 2 (PAK-2 appears to have a dual function in the regulation of cell survival and cell death. Activation of full-length PAK-2 by the p21 G-proteins Rac or Cdc42 stimulates cell survival. However, PAK-2 is unique among the PAK family because it is also activated through proteolytic cleavage by caspase 3 or similar caspases to generate the constitutively active PAK-2p34 fragment. Caspase activation of PAK-2 correlates with the induction of apoptosis in response to many stimuli and recombinant expression of PAK-2p34 has been shown to stimulate apoptosis in several human cell lines. Here, we show that caspase activation of PAK-2 also occurs during cisplatin-induced apoptosis of SH-SY5Y neuroblastoma cells as well as in SH-SY5Y cell culture models for Alzheimer’s and Parkinson’s disease. Inhibition of mitochondrial complex I or of ubiquitin/proteasome-mediated protein degradation, which both appear to be involved in Parkinson’s disease, induce apoptosis and caspase activation of PAK-2 in SH-SY5Y cells. Overexpression of the amyloid precursor protein, which results in accumulation and aggregation of β-amyloid peptide, the main component of β-amyloid plaques in Alzheimer’s disease, also induces apoptosis and caspase activation of PAK-2 in SH-SY5Y cells. Expression of the PAK-2 regulatory domain inhibits caspase-activated PAK-2p34 and prevents apoptosis in 293T human embryonic kidney cells, indicating that caspase activation of PAK-2 is directly involved in the apoptotic response. This is the first evidence that caspase activation of PAK-2 correlates with apoptosis in cell culture models of Alzheimer’s and Parkinson’s disease and that selective inhibition of caspase-activated PAK-2p34 could prevent apoptosis.

  4. Verification of the CASMO-3/SIMULATE-3 pin power accuracy by comparison with operating boiling water reactor measurements

    International Nuclear Information System (INIS)

    Uegata, T.; Saji, E.; Tanaka, H.

    1993-01-01

    Intranodal pin power distributions calculated by the CASMO-3/SIMULATE-3 code have been compared with pin gamma scan measurements. These data were obtained from the depleted core of an operating boiling water reactor (BWR), which is more complicated than a pressurized water reactor to calculate because of the existence of coolant void distributions and cruciform control blades. Furthermore, measured bundles include mixed-oxide (MOX) bundles in which steep thermal flux gradients occur. Both UO 2 and MOX bundles have been calculated in the same manner based on the standard CASMO-3/SIMULATE-3 methods. The total pin power root-mean-square (rms) error is 2.7%, which includes measurement error, from an 896-point comparison. There is no obvious dependency on axial elevations (void fractions) and no significant difference between fuel types (UO 2 or MOX), although the errors in a peripheral bundle, which is less important from the standpoint of core design, are somewhat larger than those in the internal bundles. If the peripheral bundle is excluded, the total rms error is reduced to 2.2%. From these results, it is concluded that excellent agreement has been obtained between the calculations and measurements and that the calculational capability of CASMO-3/SIMULATE-3 for the intranodal pin power distribution is quite satisfactory and useful for BWR core design

  5. Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

    International Nuclear Information System (INIS)

    2014-02-01

    JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)

  6. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    Dupraz, B.; Bertel, E.

    2003-01-01

    The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

  7. Chernobyl lesson

    Energy Technology Data Exchange (ETDEWEB)

    Vajda, G

    1986-01-01

    Structure and major technological parameters of the RBMK-1000 type Chernobylsk reactor, description of different phases of the reactor accident, the causes and consequences of the catastrophe and the measures taken to cease the fire, to stop the chain reaction, to prevent the inhabitants and the environment from radiation exposure and contamination are discussed. Major development projects at the Paks Nuclear Power Plant to support human control activities and to increase the operational safety are listed. (V.N.). 2 refs.

  8. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

  9. Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1993-01-01

    Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

  10. Experimental investigation and modelling of tritium washout by precipitation in the area of the nuclear power plant of Paks, Hungary.

    Science.gov (United States)

    Köllo, Z; Palcsu, L; Major, Z; Papp, L; Molnár, M; Ranga, T; Dombóvári, P; Manga, L

    2011-01-01

    Tritium occurs in nature in trace amounts, but its concentration is changing due to natural and artificial sources. Studies focusing on natural tritium have to take into account the effect of artificial sources. Also, the impact of tritium is an important issue in environmental protection, e.g. in connection with the emissions from nuclear power plants. The present work focuses on the rain washout of tritium emitted from the Paks nuclear power plant in Hungary. Rainwater collectors were placed around the plant and after a period of precipitation, rainwater was collected and analysed for tritium content. Samples were analysed using low-level liquid scintillation counting, with some also subject to the more accurate (3)He ingrowth method. The results clearly show the trace of the tritium plume emitted from the plant; however, values are only about one order of magnitude higher than environmental background levels. A washout model was devised to estimate the distribution of tritium around the plant. The model gives slightly higher concentrations than those measured in the field, but in general the agreement is satisfactory. The modelled values demonstrate that the effect of the plant on rainwater tritium levels is negligible over a distance of some kilometres. Copyright © 2010 Elsevier Ltd. All rights reserved.

  11. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  12. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  13. Compact stellarators as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

    2001-01-01

    Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

  14. Radiation protection at the RA Reactor in 1993, RA research reactor, Part

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

    1993-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry and radiation protection at the RA reactor; (2) decontamination, collecting and treatment of fluid effluents and solid wastes; (3) Radioactivity control in the vicinity of the reactor and (4)meteorology measurements; (3). Each of the category is described as a separate annex of this report [sr

  15. Corrosion study of heat exchanger tubes in pressurized water cooled nuclear reactors by conversion electron Moessbauer spectroscopy

    International Nuclear Information System (INIS)

    Homonnay, Z.; Kuzmann, E.; Varga, K.; Nemeth, Z.; Szabo, A.; Rado, K.; Schunk, J.; Tilky, P.; Patek, G.

    2005-01-01

    Nuclear energy production tends to return into the focus of interest because of the constantly increasing energy need of the world and the green house effect problems of the strongest competitor oil or gas based power plants. In addition to the construction of new nuclear power plants, lifetime extension of the existing ones is the most cost effective investment in the energy business. However, feasibility and safety issues become very important at this point, and corrosion of the construction materials should be carefully investigated before decision on a potential lifetime extension of a reactor. 57 Fe-Conversion Electron Moessbauer Spectroscopy (CEMS) is a sensitive tool to analyze the phase composition of corrosion products on the surface of stainless steel. The upper ∼300 nm can be investigated due to the penetration range of conversion electrons. The corrosion state of heat exchanger tubes from the four reactor units of the Paks Nuclear Power Plant, Hungary, were analyzed by several methods including CEMS. The primary circuit side of the tubes was studied on selected samples cut out from the heat exchangers during regular maintenance. Cr- and Ni-substituted magnetite, sometimes hematite, amorphous Fe-oxides/oxyhydroxides as well as the signal of bulk austenitic steel of the tubes were detected. The level of Cr- and Ni-substitution in the magnetite phase could be estimated from the Moessbauer spectra. Correlation between earlier decontamination cycles and the corrosion state of the heat exchangers was sought. In combination with other methods, a hybrid structure of the surface oxide layer of several microns was established. It is suggested that previous AP-CITROX decontamination cycles can be responsible for this structure which makes the oxide layer mobile. This mobility may be responsible for unwanted corrosion product transport into the reactor vessel by the primary coolant.

  16. A successful approach for the implementation of symptom-based emergency operating procedures for VVER reactors

    International Nuclear Information System (INIS)

    Lhoest, V.; Prior, R.; Pascal, G.

    2000-01-01

    The paper provides an overview of the organization, the progress and the results of the various Emergence Operating Procedure (EOP) development programs for VVER type reactors conducted by Westinghouse so far. The detailed working process is presented through the solutions to some major plant issues. The EOPs have been developed for the Temelin, Dukovany, Bohunice, Mochovce and Paks VVER nuclear power plants. The procedures are developed in working teams of experts from the utility and Westinghouse. The completion of the programs constitute an indication of the overall success of this approach. This is further reinforced by the general acceptance of the new procedures by the plant personnel, together with the good results obtained so far from procedure testing. This is also confirmed by a new PSA-level 1 analysis for Dukovany plant, which shows a significant improvement in the overall plant safety. This means a 20% reduction in the Core Damage Frequency due to the introduction of the new EOPs. The fact that some modifications have been implemented to the plants to solve design weaknesses identified in the course of this programs also constitute a positive result

  17. Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

    2004-01-01

    The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

  18. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  19. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program

    International Nuclear Information System (INIS)

    Hernandez, N.; Alonso, G.; Valle, E. del

    2004-01-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  20. Survey of group data libraries for use of the DYN3D program for WWER type reactors

    International Nuclear Information System (INIS)

    Mittag, S.

    1994-06-01

    So-called few-group neutron data have to be used as input data in core models (such as DYN3D) calculating the reactor behaviour. A survey is given of qualified data libraries for the reactor cores of Russian VVER. The information about primary data used in group data generation and the accuracy reached by the cell codes is compiled in tables. To assess the quality of the data, comparisons have been made between measured and calculated reactor parameters. The information available does not show significant differences concerning the quality of the data libraries. (orig.) [de