WorldWideScience

Sample records for reaktoru typu vver

  1. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  2. Fotokatalytická inaktivace kvasinek v průtočném reaktoru

    OpenAIRE

    Lipenská, Michaela

    2008-01-01

    Tato diplomová práce se zabývá fotokatalytickým účinkem oxidu titaničitého a UV záření. V teoretické části je nastíněn mechanismus působení oxidu titaničitého a jeho aplikace v různých odvětvích. Na kvasince Hansenula anomala byl sledován mikrobicidní účinek oxidu titaničitého ve spojení s UV zářením v průtočném reaktoru. Po zjištění vhodného průtoku byly voleny různé počáteční koncentrace buněk. Pro zvýšení účinku bylo do systému vneseno stříbro, které, jak bylo v mnoha pracech uvedeno, sniž...

  3. Corrosion product behavior in VVER secondary systems

    Energy Technology Data Exchange (ETDEWEB)

    Yurmanov, V.A.; Velikopolsky, S.V.; Yurmanov, E.V. [N.A. Dollezhal Research and Development Inst. of Power Engineering (NIKIET), Moscow (Russian Federation)

    2010-07-01

    Accumulation of corrosion products lead to some problems during long-term operation of VVER plants, such as secondary system component degradation including crud-induced local corrosion and corrosion cracking. Corrosion sludge and deposit removal from steam generators and other equipment is costly and time-consuming and leads to additional waste production. This problem is vital in the case of plant life extension. Appropriate solutions of the problem could be developed based on both Russian and international experience of the VVER fleet. Recommendations on how to mitigate corrosion product accumulation in VVER secondary systems were developed based on comparative analysis of available long-term data on corrosion product behavior in all the operating VVER plants, such as the following: Sludge and deposit accumulation in inner surfaces of secondary piping and components; Corrosion rate measurements using in-situ specimen testing at operated VVER plants; Efficiency of corrosion product removal from secondary system water by means of condensate polishers and steam generator blowdown cleanup systems; Sludge and deposit removal from steam generators during chemical cleaning; Secondary piping and components conservation efficiency during long outages. Comparative data analysis of corrosion product behavior has shown different corrosion product accumulation rates in Novovoronezh, Kola, Kalinin, Balakovo and Rostov NPPs. The said difference is due to different design and operation peculiarities. (author)

  4. The vver severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V. [Russian Research Center, Kurchatov Institute, Moscow (Russian Federation)

    2001-07-01

    The basic approach to the VVER safety management is based on the defence-in-depth principle the main idea of which is the multiplicity of physical barriers on the way of dangerous propagation on the one hand and the diversity of measures to protect each of them on the other hand. The main events of severe accident with loss of core cooling at NPP with WWER can be represented as a sequence of NPP states, in which each subsequent state is more severe than the previous one. The following sequence of states of the accident progression is supposed to be realistic and the most probable: -) loss of efficient core cooling; -) core melting, relocation of the molten core to the lower head and molten pool formation, -) reactor vessel damage, and -) containment damage and fission products release. The objectives of accident management at the design basis stage, the determining factors and appropriate determining parameters of processes are formulated in this paper. The same approach is used for the estimation of processes parameters at beyond design basis accident progression. The accident management goals and the determining factors and parameters are also listed in that case which is characterized by the loss of integrity of the fuel cladding. The accident management goal at the stage of core melt relocation implies the need for an efficient core-catcher.

  5. Wielonapięciowe Wentylatory Górnicze typu WWG firmy Eko-Win

    Directory of Open Access Journals (Sweden)

    Ryszard Krzykowski

    2015-04-01

    Full Text Available Przedstawiono konstrukcje typoszeregu wielonapięciowych wentylatorów górniczych typu WWG firmy Eko-Win, z omówieniem ich charakterystyk, parametrów technicznych oraz pola zastosowań. Przedstawiono sposób wyznaczania charakterystyk wentylatorów na stanowisku badawczym.

  6. Proizvodnja mliječne kiseline na hidrolizatu pšenične slame dobivenom alkalnom predobradom u visokotlačnom reaktoru

    OpenAIRE

    Vidović, Petra

    2017-01-01

    Otpadne lignocelulozne sirovine, u odnosu na šećerne i škrobne sirovine koje se uvelike koriste za prehranu ljudi i životinja, predstavljaju održivu alternativu za proizvodnju biokemikalija kao što je mliječna kiselina. U ovom radu provedena je predobrada pšenične slame s 2 %-tnom natijevom lužinom u visokotlačnom reaktoru pri različitim temperaturama (120°C-210°C) i vremenima zadržavanja od 1 do 20 minuta. Nakon predobrade pšenične slame, dobivene su dvije faze (čvrsta i tekuća faza) te je o...

  7. Estimation of material degradation of VVER-1000 baffle

    Science.gov (United States)

    Harutyunyan, Davit; Koš'ál, Michal; Vandlík, Stanislav; Hojná, Anna; Schulc, Martin; Flibor, Stanislav

    2017-09-01

    The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

  8. Estimation of material degradation of VVER-1000 baffle

    Directory of Open Access Journals (Sweden)

    Harutyunyan Davit

    2017-01-01

    Full Text Available The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

  9. VVER-1000 MOX Core Computational Benchmark: Specification and Results

    National Research Council Canada - National Science Library

    Mikhail Kalugin; Eugeny Gomin; Dmitry Oleynik

    2006-01-01

    This report presents the VVER MOX Core Computational Benchmark Specification and Results, which was proposed as a benchmark within the OECD/NEA Expert Group on Reactor-based Plutonium Disposition (TFRPD...

  10. KARATE - a code for VVER-440 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.

    1994-12-31

    A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.

  11. Zaostrzenie przewlekłej choroby nerek w przebiegu pierwotnej hiperoksalurii typu 1 - prezentacja przypadku

    Directory of Open Access Journals (Sweden)

    Violetta Bochniewska

    2010-06-01

    Full Text Available Przewlekła choroba nerek (PChN zaliczana jest do schorzeń cywilizacyjnych XXI wieku. Jej częstość w ciągu ostatnich lat systematycznie wzrasta. Według danych epidemiologicznych na PChN choruje około 600 min ludzi na świecie i ponad 4 min osób w Polsce. Najczęstsze przyczyny przewlekłej choroby nerek to: cukrzyca, nadciśnienie tętnicze, przewlekłe kłębkowe zapalenie nerek przewlekłe odmiedniczkowe zapalenie nerek, kamica dróg moczowych, zwyrodnienie torbielowate nerek nowotwory układu moczowego, szpiczak mnogi, skrobiawica wtórna. Do rzadkich przyczyn PChN możemy zaliczyć choroby uwarunkowane genetycznie, wśród nich pierwotną hiperoksalurię typu 1. W pracy autorzy przedstawiają historię rozpoznania pierwotnej hiperoksalurii typu 1 u 4-letniej dziewczynki oraz wpływ tej choroby na rozwój przewlekłej choroby nerek. Podczas leczenia ostrego odmiedniczkowego zapalenia nerek u pacjentki rozpoznano niewydolność nerek, nefrokalcynozę oraz kamicę układu moczowego. Wdrożono leczenie urologiczne, które powtarzano przez 3 miesiące (2 x operacje, 2 x ESWL. Następnie wykonano diagnostykę metaboliczną kamicy i nefrokalcynozy, która ujawniła znacznie zwiększone wydalanie kwasu szczawiowego. Z tego powodu wysunięto podejrzenie pierwotnej hiperoksalurii, które następnie potwierdzono badaniem genetycznym (typowe mutacje genu AGXT. Wdrożenie leczenia przyczynowego pirydoksyną pozwoliło na zmniejszenie wydalania kwasu szczawiowego w moczu i poprawę funkcji nerek. Wnioski: Pierwotna hiperoksaluiia typu 1 może prowadzić do niewydolności nerek w okresie wczesnego dzieciństwa. Wczesna diagnostyka i rozpoznanie tej choroby pozwalają na wdrożenie odpowiedniego leczenia, co zwiększa szansę na opóźnienie jej postępu.

  12. Methodological studies on the VVER-440 control assembly calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hordosy, G.; Kereszturi, A.; Maraczy, C. [KFKI Atomic Energy Research Institute, Budapest (Hungary)

    1995-12-31

    The control assembly regions of VVER-440 reactors are represented by 2-group albedo matrices in the global calculations of the KARATE code system. Some methodological aspects of calculating albedo matrices with the COLA transport code are presented. Illustrations are given how these matrices depend on the relevant parameters describing the boron steel and steel regions of the control assemblies. The calculation of the response matrix for a node consisting of two parts filled with different materials is discussed.

  13. Diversification of the VVER fuel market in Eastern Europe and Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Kirst, Michael [Westinghouse EMEA, Brussels (Belgium); Benjaminsson, Ulf; Oenneby, Carina [Westinghouse Electric Sweden AB, Vaesteraes (Sweden)

    2015-03-15

    There are a total of 33 VVER active reactors in the EU and Ukraine, accounting for the largest percentage of the total electricity supply in the countries operating these. The responsible governments and utilities operating these units want too see an increased diversification of the nuclear fuel supply. Westinghouse is the only nuclear fuel producer outside Russia, which has taken the major steps to develop, qualify and manufacture VVER fuel designs - both for VVER-440 and VVER-1000 reactors. The company has delivered reloads of VVER-440 fuel to Loviisa 2 in Finland, VVER-1000 fuel for both the initial core and follow-on regions to Temelin 1-2 in the Czech Republic and more recently reloads of VVER-1000 fuel to South Ukraine 2-3. Technical challenges in form of mechanical interference with the resident fuel have been encountered in Ukraine, but innovative solutions have been developed and successfully implemented and today Ukraine has, for the first time in its history, a viable VVER-1000 fuel design alternative, representing a tremendous lever in energy security for the country.

  14. Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

    Science.gov (United States)

    Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

    2017-01-01

    This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

  15. Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: alpsordo@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2012-10-15

    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

  16. VVER Reactor Safety in Eastern Europe and Former Soviet Union

    Science.gov (United States)

    Papadopoulou, Demetra

    2012-02-01

    VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

  17. ANDREA 2.2 and 2.3. Advances in modelling of VVER cores

    Energy Technology Data Exchange (ETDEWEB)

    Havluj, Frantisek; Hejzlar, Jonatan; Vocka, Radim; Vysoudil, Jiri [UJV Rez, Husinec-Rez (Czech Republic)

    2017-09-15

    In 2016 a new version of code ANDREA for core design and reload safety analysis of VVER reactors has been released. The new code version includes several major improvements. The first of them is a seamless incorporation of short time kinetics calculations (without temperature feedback) into the code. This new feature accompanied by the possibility of excore detector signal predictions enables precise interpretation of dynamic measurements of control assembly weight during the reactor startup. Second important enhancement resides in new flexible format of cross section libraries and in new fuel temperature model based on results of TRANSURANUS fuel performance code. The new code version has been thoroughly tested and validated for both VVER440 and VVER-1000 reactors. Furthermore for the new version 2.3 which is to be released shortly we have implemented the possibility of fluent control assemblies' motion and of non-equidistant axial nodalization schemes in VVER-440 calculations.

  18. Results of Post Irradiation Examinations of VVER Leaky Rods

    Energy Technology Data Exchange (ETDEWEB)

    Markov, D.; Perepelkin, S.; Polenok, V.; Zhitelev, V.; Mayorshina, G. [Head of Fuel Research Department, JSC ' SSC RIAR' , 433510, Dimitrovgrad-10, Ulyanovsk region (Russian Federation)

    2009-06-15

    The most important requirement imposed on fuel elements is to maintain integrity of fuel rod claddings under operation, storage and transportation, since it is directly related to the operational safety. However, failed rod claddings are sometimes observed under reactor operation. Identification and unloading of fuel assemblies with leaky rods from VVER is available only at the time of planned preventive maintenance. An unscheduled reactor shutdown due to the excess of coolant activity limit as well as a preterm unloading of the fuel assembly cause economic damage to nuclear plant. Therefore, models and calculation codes were developed to forecast coolant contamination and failed fuel rod behavior. Criteria based on calculations were set to determine the admissible number of the failed rods in core and the opportunity to continue the reactor operation or pre-term unloading of the fuel assembly with the failed rods. Nevertheless, to prevent the fuel rod failure (for unfailing operation) it is necessary to reveal disadvantages of the design, fabrication method and fuel operation conditions, and to eliminate defects. The most complete and significant information about spent fuel assemblies may be received following the post irradiation material examinations. In order to reveal failure origins and mechanism of changes in VVER fuel and failed rod cladding condition depending on the operation, the examinations of 12 VVER-1000 fuel assemblies and 3 VVER-440 fuel assemblies, operated under normal conditions up to the fuel burnup 13..47 MWd/kgU were carried out. To evaluate the rod cladding condition, reveal defects and determine their parameters, the ultrasonic control of cladding integrity, surface visual inspection, eddy current defectoscopy, measurement of geometrical parameters were applied. In separate cases we used the metallography, measured the hydrogen percentage and carried out the mechanical tests of o-ring samples. The pellet condition was evaluated in

  19. Návrh číslicového filtru typu pásmová propust

    OpenAIRE

    Dvořák, Vojtěch

    2011-01-01

    Cílem práce je vysvětlit problematiku digitálních filtrů IIR, ukázat postup návrhu digitálního filtru v prostředí Matlab a navrhnout model ideálního filtru typu pásmová propust s konkrétními parametry v prostředí Matlab. Tento filtr bude následně sloužit jako referenční model pro verifikaci s filtrem popsaným v jazyce VHDL. The aim of this work is to explain the problems of digital IIR filters, demostrate process of designing digital filters in Matlab and design a model of ideal band-pass ...

  20. Investigation of station blackout scenario in VVER440/v230 with RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Gencheva, Rositsa Veselinova, E-mail: roseh@mail.bg; Stefanova, Antoaneta Emilova, E-mail: antoanet@inrne.bas.bg; Groudev, Pavlin Petkov, E-mail: pavlinpg@inrne.bas.bg

    2015-12-15

    Highlights: • We have modeled SBO in VVER440. • RELAP5/MOD3 computer code has been used. • Base case calculation has been done. • Fail case calculation has been done. • Operator and alternative operator actions have been investigated. - Abstract: During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

  1. VVER Knowledge Preservation and Transfer within the Frame of CORONA Project Activities

    Science.gov (United States)

    Mitev, Mladen; Corniani, Enrico; Manolova, Maria; Pironkov, Lybomir; Tsvetkov, Iskren

    2016-02-01

    The CORONA project is funded by the European Commission under the FP7 programme with overall objective to establish a Regional Centre of Competence for VVER Technology and Nuclear Applications. The Centre will provide support and services for preservation and transfer of VVER-related nuclear knowledge as well as know-how and capacity building. Specific training schemes aimed at nuclear professionals and researchers, non-nuclear professionals and students are developed and implemented in cooperation with local, national and international training and educational institutions. Pilot trainings are executed for each specific target group to assess the applicability of the training materials. The training scheme implemented for nuclear professionals and researchers is focussed on the NPP Lifetime Management. The available knowledge on enhancing safety and performance of nuclear installations with VVER technology is used in the preparation of the training materials. The Online Multimedia Training Course on VVER Reactor Pressure Vessel Embrittlement and Integrity Assessment, developed by the joint effort of JRC-IET and IAEA is used in the training. The outcome collected from the trainees showed that the tool meets its primary goal of consolidating the existing knowledge on the VVER RPV Embrittlement and Integrity Assessment, provides adequate ground for transfer of this knowledge.

  2. Development and validation of coupled dynamics code 'TRIKIN' for VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Obaidurrahman, K; Doshi, J. B.; Jain, R. P. [IIT Bombay, Mumbai (India); Jagannathan, V. [Bhabha Atomic Research Centre, Mumbai (India)

    2010-06-15

    New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors

  3. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  4. Vliv léčby kyselinou acetylosalycilovou na makrovaskulární reaktivitu u diabetiků 2.typu

    Czech Academy of Sciences Publication Activity Database

    Prázdný, M.; Hrach, Karel; Kasalová, Z.; Škrha, J.; Zvárová, Jana

    2005-01-01

    Roč. 8, 1 Suppl. (2005), s. 41 ISSN 1211-9326. [Diabetologické dny. 21.04.2005-23.04.2005, Luhačovice] R&D Projects: GA MŠk LN00B107 Institutional research plan: CEZ:AV0Z10300504 Keywords : diabetes mellitus 2. typu * mikrovaskulární reaktivita Subject RIV: BB - Applied Statistics, Operational Research

  5. Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L.; Kaplar, E.; Lioutov, K. [Nuclear Safety Inst. of Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.; Smirnov, V.; Prokhorov, V.; Goryachev, A. [State Research Centre, Dimitrovgrad (Russian Federation). Research Inst. of Atomic Reactors

    1998-03-01

    Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in this paper.

  6. Zastosowanie termografii w diagnostyce wielomiejscowego zespołu odruchowego typu I u 42-letniego chorego

    Directory of Open Access Journals (Sweden)

    Józef Mróz

    2014-03-01

    Full Text Available Wielomiejscowy zespół odruchowy typu I jest jednostką chorobową o nie w pełni poznanym patomechanizmie i przebiegu. Charakteryzuje się silnym bólem dystalnej części kończyny, obrzękiem, dysfunkcją naczynioruchową i upośledzoną sprawnością. Objawy te występują po urazach, operacjach na klatce piersiowej, po zawale serca, po udarze, uszkodzeniu nerwów obwodowych, rzadziej w przebiegu zakrzepicy żylnej lub tętniczej. Mechanizm powstawania choroby jest niejasny. Na pierwszy plan wysuwają się zaburzenia funkcji autonomicznego układu nerwowego. Bólowi i obrzękowi kończyny w przypadkach o typowym przebiegu towarzyszą: zaburzenia naczynioruchowe, ograniczona bólem ruchomość oraz wzmożona wrażliwość na ucisk i zmiany temperatury otoczenia. W przebiegu zespołu wyróżnia się trzy okresy: I – ostry, II – dystroficzny, III – atroficzny. Poza postacią odruchową zespołu występuje postać porażenna (po udarze i toksyczna (polekowa. U większości chorych zwracają uwagę labilność emocjonalna, hiperreaktywność oraz tendencja do stanów lękowych i depresji. Przydatne w rozpoznaniu są badania obrazowe. Skuteczność leczenia zależy od okresu, w którym ustalono rozpoznanie i podjęto terapię. W leczeniu farmakologicznym stosuje się leki przeciwbólowe oraz hamujące układ współczulny. Dobre efekty przeciwobrzękowe i przeciwzapalne, a także stymulację uwapnienia kości uzyskuje się po zastosowaniu: zmiennego pola magnetycznego niskiej częstotliwości, lasera niskoenergetycznego, masażu wirowego kończyn, ćwiczeń indywidualnych zajętej kończyny. W pracy przedstawiono przypadek 42-letniego mężczyzny z wielomiejscowym zespołem odruchowym typu I, u którego diagnozę postawiono dopiero w zaawansowanym II okresie choroby. Zastosowane leczenie farmakologiczne i fizykoterapeutyczne po 3 miesiącach doprowadziło do znaczącej poprawy, którą dobrze obrazuje badanie termograficzne.

  7. Simulation of a nuclear accident by an academic simulator of a VVER-1000 reactor; Simulacion de un accidente nuclear, mediante un simulador academico de un reactor VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez G, L. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Salazar S, E., E-mail: laurahg42@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2014-10-15

    This work is planned to simulate a scenario in which the same conditions that caused the accident at the Fukushima Daichi nuclear power plant are present, using a simulator of a nuclear power plant with VVER-1000 reactor, a different type of technology to the NPP where the accident occurred, which used BWR reactors. The software where it will take place the simulation was created and distributed by the IAEA for academic purposes, which contains the essential systems that characterize this type of NPP. The simulator has tools for the analysis of the characteristic phenomena of a VVER-1000 reactor in the different systems together and planned training tasks. This makes possible to identify the function of each component and how connects to other systems, thus facilitating the visualization of possible failures and the consequences that they have on the general behavior of the reactor. To program the conditions in the simulator, is necessary to know and synthesize a series of events occurred in Fukushima in 2011 and the realized maneuvers to reduce the effects of the system failures. Being different technologies interpretation of the changes that would suffer the VVER systems in the scenario in question will be developed. The Fukushima accident was characterized by the power loss of regular supply and emergency of the cooling systems which resulted in an increase in reactor temperature and subsequent fusion of their nuclei. Is interesting to reproduce this type of failure in a VVER, and extrapolate the lack of power supply in the systems that comprise, as well as pumping systems for cooling, has a pressure regulating system which involves more variables in the balance of the system. (Author)

  8. The corrosion and corrosion mechanical properties evaluation for the LBB concept in VVERs

    Energy Technology Data Exchange (ETDEWEB)

    Ruscak, M.; Chvatal, P.; Karnik, D.

    1997-04-01

    One of the conditions required for Leak Before Break application is the verification that the influence of corrosion environment on the material of the component can be neglected. Both the general corrosion and/or the initiation and, growth of corrosion-mechanical cracks must not cause the degradation. The primary piping in the VVER nuclear power plant is made from austenitic steels (VVER 440) and low alloy steels protected with the austenitic cladding (VVER 1000). Inspection of the base metal and heterogeneous weldments from the VVER 440 showed that the crack growth rates are below 10 m/s if a low oxygen level is kept in the primary environment. No intergranular cracking was observed in low and high oxygen water after any type of testing, with constant or periodic loading. In the framework of the LBB assessment of the VVER 1000, the corrosion and corrosion mechanical properties were also evaluated. The corrosion and corrosion mechanical testing was oriented predominantly to three types of tests: stress corrosion cracking tests corrosion fatigue tests evaluation of the resistance against corrosion damage. In this paper, the methods used for these tests are described and the materials are compared from the point of view of response on static and periodic mechanical stress on the low alloyed steel 10GN2WA and weld metal exposed in the primary circuit environment. The slow strain rate tests and static loading of both C-rings and CT specimens were performed in order to assess the stress corrosion cracking characteristics. Cyclic loading of CT specimens was done to evaluate the kinetics of the crack growth under periodical loading. Results are shown to illustrate the approaches used. The data obtained were evaluated also from the point of view of comparison of the influence of different structure on the stress corrosion cracking appearance. The results obtained for the base metal and weld metal of the piping are presented here.

  9. Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Del Nevo, A., E-mail: alessandro.delnevo@enea.it [ENEA, C.R. Brasimone, 40032 Camugnano (Italy); Moretti, F.; D' Auria, F. [GRNSPG, Universita di Pisa, via Diotisalvi 2, 56100 Pisa (Italy); Elkin, I.V.; Melikhov, O.I. [Electrogorsk Research and Engineering Centre, Electrogorsk, Moscow Region (Russian Federation)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Four integral test facility experiments related to VVER-1000 reactor. Black-Right-Pointing-Pointer TH response of the VVER-1000 primary system following total loss of feedwater and station blackout scenarios. Black-Right-Pointing-Pointer Accident management procedures in case of total loss of feedwater and station blackout. Black-Right-Pointing-Pointer Experimental data represent an improvement of existing database for TH code validation. - Abstract: VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project 'TACIS 2.03/97', Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.

  10. EVALUATION METRICS APPLIED TO ACCIDENT TOLERANT FUEL CLADDING CONCEPTS FOR VVER REACTORS

    Directory of Open Access Journals (Sweden)

    Martin Sevecek

    2016-12-01

    Full Text Available Enhancing the accident tolerance of LWRs became a topic of high interest in many countries after the accidents at Fukushima-Daiichi. Fuel systems that can tolerate a severe accident for a longer time period are referred as Accident Tolerant Fuels (ATF. Development of a new ATF fuel system requires evaluation, characterization and prioritization since many concepts have been investigated during the first development phase. For that reason, evaluation metrics have to be defined, constraints and attributes of each ATF concept have to be studied and finally rating of concepts presented. This paper summarizes evaluation metrics for ATF cladding with a focus on VVER reactor types. Fundamental attributes and evaluation baseline was defined together with illustrative scenarios of severe accidents for modeling purposes and differences between PWR design and VVER design.

  11. Advanced power plant training simulator for VVER-440/V230 nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Shier, W.; Kennett, R. [Brookhaven National Lab., Upton, NY (United States); Vaclav, E.; Gieci, A. [Nuclear Power Research Inst. Trnava, Inc. (Slovakia)

    1996-11-01

    An advanced, workstation based, nuclear power plant simulator has been developed for use in training the operational staff of the Bohunice Nuclear Power Plant. This training simulator uses state-of- the-art computer hardware and software and provides the capability to simultaneously include six members of the power plant operating staff in the training sessions. A detailed reactor model has been developed, representing the Bohunice VVER-44O/V230 plants, for use with the RELAP5 simulation software. In addition, a comprehensive validation program has been completed that compares the simulation results of the advanced simulator with the results from a current VVER-44O/V230 simulator. A summary of the training features and capabilities of the simulator is also provided.

  12. Steam Line Break investigation at full power reactor for VVER-1000/V320

    Energy Technology Data Exchange (ETDEWEB)

    Pavlova, M., E-mail: pavlovamp@mail.bg; Andreeva, M., E-mail: m_andreeva@inrne.bas.bg; Groudev, P., E-mail: pavlinpg@inrne.bas.bg

    2015-04-15

    Highlights: • In this study we investigated Steam Line Break accident at full power reactor. • The reference power plant for the analyses is Unit 6 at Kozloduy NPP. • The RELAP/MOD 3.2 computer code is used in performing the analyses. • The results are used for analytical validation of EOP. - Abstract: This paper presents the results of thermal-hydraulic calculation of “Steam Line Break” analysis at full power reactor for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of symptom based emergency operating procedures (SB EOPs) for this plant. The RELAP5/MOD 3.2 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the systems thermal-hydraulics code RELAP5/MOD 3.2 at the Institute for Nuclear Research and Nuclear Energy–Bulgarian Academy of Sciences (INRNE–BAS), Sofia. The main purpose of the analysis is to estimate the parameters of the monitored plant which are used to identify symptoms that are used by operators to identify the plant's state and the critical safety function (CSF). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The performed analysis is based on a previously used bounding approach in analytical validation of SB EOPs. Based on this approach a list of scenarios has been performed, involving a different number of safety systems with or without operator actions. The presented thermal-hydraulic calculations of the accident scenarios involve the loss of CSF “Subcriticality” for VVER-1000/V320 units at KNPP.

  13. Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Jakubec Jakub

    2017-04-01

    Full Text Available The paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

  14. High energy pipe line break postulations and their mitigation - examples for VVER nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L.; Kadecka, P.; Dotrel, J. [Nuclear Res. Inst., Rez (Czech Republic)

    1998-11-01

    The concept and the proposals for the protection and reinforcement of equipment against the effects of postulated rupture of the high-energy piping, in VVER Plant, are presented. The most recent version of the US NRC Guidelines has been used. The development of the legislation, the basic approach and selection of criteria for the assessment of the rupture of high energy piping, provide the basis for the application of the separation concept in the overall safety philosophy. (orig.)

  15. Addressing the scaling issue with Cathare 2 simulation of VVER 1000 transient scenario

    Energy Technology Data Exchange (ETDEWEB)

    Dino Araneo; Alessandro Del Nevo; Francesco D' Auria; Giorgio Galassi [DIMNP Universita di Pisa, Via Diotisalvi, 2, 56122 Pisa (Italy)

    2005-07-01

    Full text of publication follows: Tests performed at scaled facilities play an important role in the assessment of safety of Nuclear Power Plant (NPP). The results obtained by the tests performed in the facilities can be used to qualifies the NPP nodalization. Starting from the same initial and boundary conditions of the experimental tests performed at the facility the full plant nodalization must reproduce the same phenomena with the same timing. This is indicated as 'Kv scaled calculation'. A good agreement between the results obtained in the calculation and the experimental tests allows to say that the plant nodalization is able to reproduce the behaviour of the plant in transient scenarios. This paper deals with the scaling issue concerning a Cathare2 simulation of a VVER 1000 transient scenario. The PSB-VVER facility is built in 1998 at Electrogorsk Research and Engineering Centre. It is a facility with a scaling factor of 1/300 for the volume of the referred NPP (VVER-1000). In order to evaluate the nodalization performance the qualification procedure developed at the DIMNP of Pisa University (UNIPI) has been applied. This procedure foresees two levels of qualification: a 'steady state' level and an 'on transient' level. After the steady state results of the nodalization has been checked, the 'on transient' qualification check is performed adopting the PSB-VVER 11% equivalent break in Upper Plenum. This test includes the actuation of the HPIS injecting only in the loop 4 and the availability of the hydro accumulators. (authors)

  16. Influence of geometrical parameters of the VVER-1000 reactor construction elements to internals irradiation conditions

    Directory of Open Access Journals (Sweden)

    О. M. Pugach

    2015-07-01

    Full Text Available Investigations to determine the influences of geometrical parameters of the calculational VVER-1000 reactor model to the results of internal irradiation condition determination are carried out. It is shown that the values of appropriate sensitivity matrix elements are not dependent on a height coordinate for any core level, but there is their azimuthal dependence. Maximum possible relative biases of neutron fluence due to inexact knowledge of internal geometrical parameters are obtained for the baffle and the barrel.

  17. Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Maltsev, D. A.; Fedotova, S. V.; Frolov, A. S.; Zhuchkov, G. M.

    2017-01-01

    This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (TK) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in TK shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the TK shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime.

  18. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Hennion, A

    1999-03-15

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  19. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  20. Metformina – mechanizmy działania i zastosowanie w terapii cukrzycy typu 2[i][/i

    Directory of Open Access Journals (Sweden)

    Marzena Grzybowska

    2011-01-01

    Full Text Available Metformina jest obecnie najczęściej zalecanym lekiem w terapii cukrzycy typu 2. Mimo iż ta pochodna biguanidu jest stosowana od ponad 50 lat, mechanizm jej działania nie został dokładnie poznany. W pracy przedstawiono najnowsze doniesienia o mechanizmach antyhiperglikemicznego działania metforminy. Obejmują one: zmniejszenie wchłaniania glukozy w jelicie cienkim, zwiększony transport glukozy do komórek, obniżenie osoczowego stężenia wolnych kwasów tłuszczowych oraz hamowanie glukoneogenezy. Szczególną rolę w tych procesach odgrywa aktywacja kinazy białkowej aktywowanej przez AMP. Najnowsze odkrycia umożliwiły poznanie mechanizmów działania przeciwmiażdżycowego, hipotensyjnego i przeciwnowotworowego metforminy oraz jej wpływu na czynność śródbłonka naczyń. Plejotropowe działanie metforminy obejmuje wpływ na profil lipidowy osocza, zmniejszenie stresu oksydacyjnego, a także zwiększenie aktywności fibrynolitycznej osocza. Mimo że metformina nie jest metabolizowana, najnowsze badania wykazały, że jest aktywnie transportowana do hepatocytów, a także do komórek nabłonka kanalików nerkowych, odpowiednio przez OCT1 (organic cation transporter 1, kodowany przez gen SLC22A1 oraz OCT2 (kodowany przez [i]SLC22A2[/i]. Z kolei transporter MATE1 (multidrug and toxin extrusion 1 protein, kodowany przez gen [i]SLC47A1[/i] umożliwia wydzielanie metforminy z tych komórek do żółci lub moczu. Polimorfizm genów transporterów metforminy może się przyczynić do istotnych różnic w reakcji na lek.

  1. VVER-1000 MOX Core Computational Benchmark analysis using indigenous codes EXCEL, TRIHEX-FA and HEXPIN

    Energy Technology Data Exchange (ETDEWEB)

    Thilagam, L. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)], E-mail: thilagam@igcar.gov.in; Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)], E-mail: v_jagan1952@rediffmail.com; Sunil sunny, C.; Subbaiah, K.V. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)

    2009-10-15

    Validation studies based on the analysis of theoretical benchmarks play a key role in the identification of deficiencies in the reactor physics design computational codes and the associated nuclear data libraries. Implementation of improvements, if any, in theoretical models and the choice of appropriate nuclear data libraries help in enhancing the accuracy of calculations. As part of the effort for the validation of computer codes for plutonium utilization in VVER type reactors, the indigenous codes EXCEL, TRIHEX-FA and HEXPIN, developed at Light Water Reactor Physics Section (LWRPS), RPDD, BARC, and the associated nuclear data library (JEF22XS), were employed to analyse 'VVER-1000 MOX Core Computational Benchmark'. The few group homogenized parameters of assembly cell or individual lattice cells were obtained by the hexagonal lattice burn-up code EXCEL and the core diffusion calculations were then performed using hexagonal assembly geometric code TRIHEX-FA or the pin-by-pin diffusion code HEXPIN. VVER-1000 reactor core loaded with 2/3rd of Low-Enriched Uranium (LEU) fuel assemblies (FAs) and 1/3rd of weapons grade MOX FAs was investigated. Effective multiplication factors and assembly average fission reaction rate distributions have been calculated for various reactor state descriptions using 3-D diffusion theory codes TRIHEX-FA and HEXPIN. Further, estimate of detailed pin-by-pin fission reaction rate distributions of a few selected assemblies were made for the normal working state of the reactor using pin-by-pin core simulation code HEXPIN. A comparison of results was done with the reported Monte Carlo (MC) values of the benchmark and in most cases good agreement was observed with the benchmark results.

  2. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1995-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  3. SEARCH FOR THE BEST POWER CONTROL PROGRAM AT NPP WITH VVER-1000 USING GRADIENT DESCENT METHOD

    Directory of Open Access Journals (Sweden)

    S. N. Pelykh

    2016-09-01

    Full Text Available This article is regarded to the search for the best power control program at nuclear power plant (NPP with VVER- 1000 by gradient descent method for the objective function, which includes the criteria of efficiency, safety and damage. Criteria normalization to the maximum value is carried out when looking for the minimum of the objective function because criteria have different physical nature. There were chosen such objective criteria as depth of fuel burn-up, index of the fuel cladding damage and axial offset - the ratio of the energy at the top and bottom of the reactor core.

  4. Department of Energy's team's analyses of Soviet designed VVERs (water-cooled water-moderated atomic energy reactors)

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document contains apprendices A through P of this report. Topics discussed are: a cronyms and technical terms, accident analyses reactivity control; Soviet safety regulations; radionuclide inventory; decay heat; operations and maintenance; steam supply system; concrete and concrete structures; seismicity; site information; neutronic parameters; loss of electric power; diesel generator reliability; Soviet codes and standards; and comparisons of PWR and VVER features. (FI)

  5. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.

    2000-06-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  6. CFD Analysis of a Slug Mixing Experiment Conducted on a VVER-1000 Model

    Directory of Open Access Journals (Sweden)

    F. Moretti

    2009-01-01

    Full Text Available A commercial CFD code was applied, for validation purposes, to the simulation of a slug mixing experiment carried out at OKB “Gidropress” scaled facility in the framework of EC TACIS project R2.02/02: “Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration at core inlet.” Such experimental model reproduces a VVER-1000 nuclear reactor and is aimed at investigating the in-vessel mixing phenomena. The addressed experiment involves the start-up of one of the four reactor coolant pumps (the other three remaining idle, and the presence of a tracer slug on the starting loop, which is thus transported to the reactor pressure vessel where it mixes with the clear water. Such conditions may occur in a boron dilution scenario, hence the relevance of the addressed phenomena for nuclear reactor safety. Both a pretest and a posttest CFD simulations of the mentioned experiment were performed, which differ in the definition of the boundary conditions (based either on nominal quantities or on measured quantities, resp.. The numerical results are qualitatively and quantitatively analyzed and compared against the measured data in terms of space and time tracer distribution at the core inlet. The improvement of the results due to the optimization of the boundary conditions is evidenced, and a quantification of the simulation accuracy is proposed.

  7. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  8. Physical startup tests for VVER-1200 of Novovoronezh NPP. Advanced technique and some results

    Energy Technology Data Exchange (ETDEWEB)

    Afanasiev, Dmitry A.; Kraynov, Yury A.; Pinegin, Anatoly A.; Tsyganov, Sergey V. [National Research Centre, Moscow (Russian Federation). Kurchatov Inst.

    2017-09-15

    The intention of the startup physics tests was to confirm design characteristics of the core loading and their compliance with safety analysis preconditions. The program of startup tests for the leading unit is usually composed in such a way that is is possible to study as much neutron-physical characteristics as possible in the safest condition of zero power. State-of-the-art safety analysis is including computer codes that use three dimensional neutron kinetics and thermohydraulics models. For the substantiation of such models, for its validation and verification there is a need in reactor experiments that implementing spatially distributed transients. We based on such statements when composing hot zero power physical startup program for the new VVER-1200 unit of Novovoronezh NPP. Several tests unconventional for VVER were developed for that program. It includes measuring the worth for each of control rod groups and measuring of single rod worth from the inserted groups - test that models rod ejection event in some sense.

  9. Consistent neutron-physical and thermal-physical calculations of fuel rods of VVER type reactors

    Directory of Open Access Journals (Sweden)

    Tikhomirov Georgy

    2017-01-01

    Full Text Available For modeling the isotopic composition of fuel, and maximum temperatures at different moments of time, one can use different algorithms and codes. In connection with the development of new types of fuel assemblies and progress in computer technology, the task makes important to increase accuracy in modeling of the above characteristics of fuel assemblies during the operation. Calculations of neutronphysical characteristics of fuel rods are mainly based on models using averaged temperature, thermal conductivity factors, and heat power density. In this paper, complex approach is presented, based on modern algorithms, methods and codes to solve separate tasks of thermal conductivity, neutron transport, and nuclide transformation kinetics. It allows to perform neutron-physical and thermal-physical calculation of the reactor with detailed temperature distribution, with account of temperature-depending thermal conductivity and other characteristics. It was applied to studies of fuel cell of the VVER-1000 reactor. When developing new algorithms and programs, which should improve the accuracy of modeling the isotopic composition and maximum temperature in the fuel rod, it is necessary to have a set of test tasks for verification. The proposed approach can be used for development of such verification base for testing calculation of fuel rods of VVER type reactors

  10. Development and application of the coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER for safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lizorkin, M.; Nikonov, S. [Kurchatov Institute for Atomic Energy, Moscow (Russian Federation); Langenbuch, S.; Velkov, K. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2006-07-01

    The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modeling capability of this coupled code as well as the status of validation by benchmark activities and comparison with plant measurements are described. The paper is focused on the modeling of flow mixing in the reactor pressure vessel including its validation and the application for the safety justification of VVER plants. (authors)

  11. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  12. Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

    Directory of Open Access Journals (Sweden)

    V.I. Belozerov

    2016-12-01

    Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR. It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

  13. CATHARE Multi-1D Modeling of Coolant Mixing in VVER-1000 for RIA Analysis

    Directory of Open Access Journals (Sweden)

    I. Spasov

    2010-01-01

    Full Text Available The paper presents validation results for multichannel vessel thermal-hydraulic models in CATHARE used in coupled 3D neutronic/thermal hydraulic calculations. The mixing is modeled with cross flows governed by local pressure drops. The test cases are from the OECD VVER-1000 coolant transient benchmark (V1000CT and include asymmetric vessel flow transients and main steam line break (MSLB transients. Plant data from flow mixing experiments are available for comparison. Sufficient mesh refinement with up to 24 sectors in the vessel is considered for acceptable resolution. The results demonstrate the applicability of such validated thermal-hydraulic models to MSLB scenarios involving thermal mixing, azimuthal flow rotation, and primary pump trip. An acceptable trade-off between accuracy and computational efficiency can be obtained.

  14. Technology of repair of selected equipment in the power plant type VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Barborka, J.; Magula, V. [Welding Research Inst. (WRI), Bratislava (Slovakia)

    1998-11-01

    This article is divided in two parts: The first part is studying the effect of individual parameters by the usual and pulsed welding of 15CH2MFA steel. It can be concluded that by use of mechanized or automatic TIG process in PC position with addition of a cold wire with high nickel content the desired quality of repair welded joints of a pressure vessel of VVER 440 reactor can be achieved. Based on the results of the second laboratory study of the renovation technology applied for the rotary surfaces of pressure-tight cover and spindle of the main closing armature type DN 500 it can be concluded, that the developed technology for surfacing the sealing surfaces by TIG process with addition of a high-nickel cold wire the functional capability of the mentioned parts can be fully restored.

  15. Economical Feedback of Increasing Fuel Enrichment on Electricity Cost for VVER-1000

    Directory of Open Access Journals (Sweden)

    Mohammed Saad Dwiddar

    2015-08-01

    Full Text Available A methodology of evaluating the economics of the front-end nuclear fuel cycle with a price change sensitivity analysis for a VVER-1000 reactor core as a case study is presented. The effect of increasing the fuel enrichment and its corresponding reactor cycle length on the energy cost is investigated. The enrichment component was found to represent the highly expenses dynamic component affecting the economics of the front-end fuel cycle. Nevertheless, the increase of the fuel enrichment will increase the reactor cycle length, which will have a positive feedback on the electricity generation cost (cent/KWh. A long reactor operation time with a cheaper energy cost set the nuclear energy as a competitive alternative when compared with other energy sources.

  16. Basket criticality design of a dual purpose cask for VVER 1000 spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rezaeian, Mahdi [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Kamali, Jamshid [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

    2016-12-15

    Dual purpose cask technology is one of the most prominent options for interim storage of spent fuels following their removal from reactors. Criticality safety of the spent fuel assemblies are ensured by design of the basket within these casks. In this study, a set of criticality design calculations of a dual purpose cask for 12 VVER 1000 spent fuel assemblies of Bushehr nuclear power plant were carried out. The basket material of borated stainless steel with 0.5 to 2.5 wt% of boron and Boral (Al-B{sub 4}C) with 1.5 to 40 wt% of boron carbide, were investigated and the minimum required receptacle pitch of the basket was determined. Using the calculated receptacle pitch of the basket, the minimum required diameter of the cavity could be established.

  17. A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL

    Energy Technology Data Exchange (ETDEWEB)

    Thilagam, L. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)], E-mail: thilagam@igcar.gov.in; Sunil Sunny, C. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India); Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)], E-mail: v_jagan1952@rediffmail.com; Subbaiah, K.V. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)

    2009-05-01

    Utilization of Mixed Uranium-Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with intent to investigate the core physics behaviour of a VVER-1000 reactor loaded with 2/3rd of low enriched uranium (LEU) fuel assemblies (FA) and 1/3rd of weapons grade mixed oxide (MOX) FA. In the present study an attempt is made to analyse 'AVVER-1000LEUandMOXAssemblyComputationalBenchmark' and predict the neutronics behaviour at the lattice level. The lattice burnup code EXCEL, developed at Light Water Reactor Physics Section, BARC is employed for this task. The EXCEL code uses the 172 energy group 'JEFF31GX' cross-section library in WIMS-D format. Assembly level fuel depletion calculations are performed up to a burnup of 40 MWD/kg of heavy metal (HM). Studies are made for the parametric variations of fuel and moderator temperatures, coolant density and boron content in the coolant. Both operational and off-normal states are analysed to determine the corresponding infinite neutron multiplication factor (k{sub {infinity}}). Pin wise isotopic compositions are computed as a function of burnup. Isotopic compositions in different annular regions of Uranium-Gadolinium (UGD) pin, fission rate distributions in UGD, UO{sub 2} and MOX pin cells are also computed. The predicted results are compared with the benchmark mean results.

  18. IMPROVED MODELS AND METHOD OF POWER CHANGE OF NPP UNIT WITH VVER-1000

    Directory of Open Access Journals (Sweden)

    Tymur Foshch

    2017-05-01

    Full Text Available This study represents the improved mathematical and imitational allocated in space multi-zone model of VVER-1000 which differs from the known one. It allows to take into account the energy release of 235U nuclei fission as well as 239Pu . Moreover, this model includes sub-models of simultaneous control impact of the boric acid concentration in the coolant of the first circuit and the position of 9th group control rods which allows to consider it as the model with allocated parameters and also allows to monitor changes in the mentioned technological parameters by reactor core symmetry sectors, by layers of reactor core height and by fuel assembly group each symmetry sector. Moreover, this model allows to calculate important process-dependent parameters of the reactor (including axial offset as quantitative measure of its safety. As the mathematical and imitational models were improved, it allows to take into account intrinsic properties of the reactor core (including transient processes of xenon and thus reduce the error of modelling static and dynamic properties of the reactor.The automated control method of power change of the NPP unit with VVER-1000 was proposed for the first time. It uses three control loops. One of which maintains the regulatory change of reactor power by regulating the concentration of boric acid in the coolant, the second circuit keeps the required value of axial offset by changing the position of control rods, and the third one holds constant the coolant temperature mode by regulating the position of the main turbo generator valves.On the basis of the above obtained method, two control programs were improved. The first one is the improved control program that implements the constant temperature of the coolant in the first circuit and the second one is the improved control program that implements the constant steam pressure in the second circuit.

  19. Sequence of decommissioning of the main equipment in a central type VVER 440 V-230; Secuencia de desmantelamiento de los equipos principales de una central Tipo VVer 440 V-230

    Energy Technology Data Exchange (ETDEWEB)

    Andres, E.; Garcia Ruiz, R.

    2014-10-01

    IBERDROLA Ingenieria y Construccion S.A.U., leader of consortium with Empresarios Agrupados and INDRA, has developed the Basic Engineering for the decommissioning of contaminated systems and building of a VVER 440 V-230 Nuclear Power Plant, establishing the sequence and methodology for the main equipment fragmentation. For that, it has been designed dry and wet cutting zones to be set up in the area where steam generators, main cooling pumps and pressurizer are located; these components will be dismantled previously. (Author)

  20. Simulation of Low-Enriched Uranium (LEU) Burnup in Russian VVER Reactors with the HELIOS Code Package

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, B.D.; Kravchenko, J.; Lazarenko, A.; Pavlovitchev, A.; Sidorenko, V.; Chetverikov, A.

    2000-03-01

    The HELIOS reactor-physics computer program system was used to simulate the burnup of UO{sub 2} fuel in three VVER reactors. The manner in which HELIOS was used in these simulations is described. Predictions of concentrations for actinides up to {sup 244}Cm and for isotopes of neodymium were compared with laboratory-measured values. Reasonable agreement between calculated and measured values was seen for experimental samples from a fuel rod in the interior of an assembly.

  1. Development of a cross-section methodology and a real-time core model for VVER-1000 simulator application

    Energy Technology Data Exchange (ETDEWEB)

    Georgieva, Emiliya Lyudmilova

    2016-06-06

    The novel academic contributions are summarized as follows. A) A cross-section modelling methodology and a cycle-specific cross-section update procedure are developed to meet fidelity requirements applicable to a cycle-specific reactor core simulation, as well as particular customer needs and practices supporting VVER-1000 operation and safety. B) A real-time version of the Nodal Expansion Method code is developed and implemented into Kozloduy 6 full-scope replica control room simulator.

  2. Analytical validation of operator actions in case of primary to secondary leakage for VVER-1000/V320

    Energy Technology Data Exchange (ETDEWEB)

    Andreeva, M., E-mail: m_andreeva@inrne.bas.bg; Groudev, P., E-mail: pavlinpg@inrne.bas.bg; Pavlova, M., E-mail: pavlova@inrne.bas.bg

    2015-12-15

    Highlights: • We validate operator actions in case of primary to secondary leakage. • We perform four scenarios related to SGTR accident for VVER-1000/V320. • The reference power plant for the analyses is Unit 6 at Kozloduy NPP. • The RELAP5/MOD 3.2 computer code is used in performing the analyses. • The analyses confirm the effectiveness of operator actions during PRISE. - Abstract: This paper presents the results of analytical validation of operator actions in case of “Steam Generator Tube Rupture” (SGTR) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The purpose of the analyses is to demonstrate the ability to terminate primary to secondary leakage and to indicate an effective strategy for preventing secondary leakage to the environment and in this way to prevent radiological release to the environment. Following depressurization and cooldown of reactor coolant system (RCS) with isolation of the affected steam generator (SG), in these analyses are validated options for post-SGTR cooldown by: • back up filling the ruptured SG; • using letdown system in the affected SG and • by opening Fast Acting Isolation Valve (FAIV) and using Steam Dump Facility to the Condenser (BRU-K). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The RELAP5/MOD3.2 computer code has been used for the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS). This paper is possible through the participation of leading specialists from KNPP.

  3. Analytical validation of operator actions based on SAMG for VVER 1000 with ASTECv2r3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, Antoaneta Emilova, E-mail: antoanet@inrne.bas.bg; Gencheva, Rositsa Veselinova, E-mail: roseh@mail.bg; Groudev, Pavlin Petkov, E-mail: pavlinpg@inrne.bas.bg

    2015-01-15

    Highlights: • Performing of analytical validation of operator action based SAMG. • Simulation of base calculation of SBO scenario without operator action for VVER 1000. • Simulation of SBO scenario with investigation of operator actions based on SAMG for VVER 1000. - Abstract: This paper presents the analytical validation of operator action based on severe accident management guidelines (SAMG) for Kozloduy NPP VVER1000 with severe accident computer code ASTECv2r3. The work is oriented on investigation of plant behavior during total loss of power and the operator actions performed based on strategies considered in severe accident management guidelines (SAMG) in Kozloduy nuclear power plant (KNPP). Using the SAMG strategies the operator depressurize primary circuit by gas removing system (YR) and try to cool down the reactor core by high pressure injection system (HPIS). The purpose of these analyses is to examine the possibility of keeping the core from further damage during a severe accident and to assess the likelihood of additional generation of hydrogen by additional flooding of the heated core. For this purpose it have been simulated a SBO scenario with injection of cold water by a high pressure pump (HPP) in cold leg at different core exit temperatures at 923 K and 1253 K. The selection of investigated analyses was based on severe accident management strategy of KNPP VVER1000. The presented work is important for analytical validation, verification, and further improvements of SAMG as well as for assessment of Level 2 probabilistic safety analyses (L2 PSA). The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research.

  4. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovitchev, A.M.

    2000-03-08

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes.

  5. The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

    Science.gov (United States)

    Arkadov, G. V.; Zhukavin, A. P.; Kroshilin, A. E.; Parshikov, I. A.; Solov'ev, S. L.; Shishov, A. V.

    2014-10-01

    The article describes the "Virtual Digital VVER-Based Nuclear Power Plant" computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the "Virtual Digital VVER-Based NPP" computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the "Virtual Digital VVER-Based NPP" computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.

  6. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

    Directory of Open Access Journals (Sweden)

    V. Sánchez

    2010-01-01

    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  7. Experimental Investigation of Operation of VVER Steam Generator in Condensation Mode in the Event of the Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, Andrey [Institute for Physics and Power Engineering by A.I. Leypunsky, 1 Bondarenko sq. Obninsk, 249033 (Russian Federation)

    2008-07-01

    For new Russian nuclear power plants with VVER-1200 reactor in the event of a beyond design basis accident, provision is made for the use of passive safety systems for necessary core cooling. These safety systems include the passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam. As a result, the condensate from SG arrives at the core providing its additional cooling. To investigate the condensation mode of VVER SG operation, a large scale HA2M-SG test facility was constructed. The rig incorporates: buffer tank, SG model with scale is 1:46, PHRS heat exchanger. Experiments at the test facility have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. The report presents the test procedure and the basic obtained test results. (authors)

  8. Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Moretti, F.; Melideo, D. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); Del Nevo, A., E-mail: delnevo@hotmail.com [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); D' Auria, F. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); Hoehne, T. [Forschungszentrum Dresden-Rossendorf (FZD), P.O.B. 51 01 19, D-01314 Dresden (Germany); Lisenkov, E. [FSUE OKB Gidropress, Ordshonikidize 21, RU-142103 Podolsk, Moscow district (Russian Federation); Gallori, D. [AREVA NP SAS, Tour AREVA - 92084 Paris, La Defense Cedex (France)

    2011-08-15

    Highlights: > Five mixing experiments in a scaled model of a VVER-1000 are described and discussed. > In-vessel mixing investigations of the coolant properties distribution at the core inlet. > These tests brought an improvement to existing experimental database for TH code validation. - Abstract: In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project 'TACIS R2.02/02 - Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet'. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.

  9. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Gy. Ézsöl

    2012-01-01

    Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

  10. Współwystępowanie nerwiakowłókniakowatości typu 1. i glejaka skrzyżowania nerwów wzrokowych – opis przypadku

    Directory of Open Access Journals (Sweden)

    Joanna Tarasiuk

    2010-03-01

    Full Text Available Nerwiakowłókniakowatość typu 1. (NF1, choroba von Recklinghausena jest zaliczana do schorzeń nerwowo-skórnych, tzw. fakomatoz. Jest to jedno z najczęstszych zaburzeń dziedziczonych autosomalnie dominująco, występujące z często- ścią około 1:3000 urodzeń. Typowe objawy NF1 to przebarwienia skórne typu „kawa z mlekiem”, piegi w okolicach pach i pachwin, nerwiakowłókniaki oraz guzki Lischa na tęczówkach. U pacjentów częściej niż w populacji ogólnej obserwuje się również występowanie zarówno nowotworów łagodnych, jak i złośliwych. Guzem najczęściej spotykanym w przypadku NF1 (tj. u około 15-20% pacjentów jest glejak drogi wzrokowej, tzw. gwiaździak włosowaty, który zajmuje nerwy wzrokowe i/lub skrzyżowanie nerwów wzrokowych oraz prowadzi do ubytków w polu widzenia. Glejaki drogi wzrokowej rosną powoli, niekiedy mogą być bezobjawowe, rzadko przybierają agresywną formę i tylko jedna trzecia pacjentów wymaga leczenia operacyjnego. Wśród innych nowotworów u chorych z NF1 spotykamy również: mięsaki nerwów obwodowych, mięsaki prążkowanokomórkowe, guz pheochromocytoma, rakowiaka dwunastnicy, białaczkę nielimfocytarną. Do innych objawów nerwiakowłókniakowatości typu 1. zaliczamy dysplazje i deformacje kostne, makrocefalię, niski wzrost, opóźnienie umysłowe, zaburzenia orientacji wzrokowo-przestrzennej, gorszą pamięć, dysleksję oraz padaczkę. W pracy przedstawiono przypadek 19-letniego chorego z rozpoznaną nerwiakowłókniakowatością typu 1. ze współistniejącym glejakiem skrzyżowania nerwów wzrokowych oraz padaczką lekooporną.

  11. The impact of ODA microadditions into secondary system on corrosion rate reduction in VVER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Avdeev, A.A.; Kukushkin, A.N.; Repin, D.A. [All-Russia Research and Design Inst. of Nuclear Power Machine Building (VNIIAM), Moscow (Russian Federation); Omelchuk, V.V.; Barmin, L.F. [Kola Nuclear Power Plant, Polyarnye Zori, Murmansk region (Russian Federation); Yurmanov, V.A. [N.A. Dollezhal Research and Development Inst. of Power Engineering (NIKIET), Moscow (Russian Federation); Czempik, E. [RECON GmbH, Leipzig (Germany)

    2010-07-01

    Injection of film-forming corrosion inhibitors is a challenging way of suppressing erosion-corrosion and crud induced corrosion in power stations. Films of surface-active inhibitors, such as octadecylamine (ODA) provide a diffusion barrier to penetration of corrosion-aggressive ions onto the metal surface. Erosion and corrosion tests were conducted in autoclaves and on a pilot steam generator (SG) design to look into the impact of ODA. To accelerate corrosion process tests were conducted in a more aggressive environment as compared to actual operating conditions, including high chloride concentration and stress levels. It is not only important to reduce deposition growth, but also to wash out deposits previously formed on heat exchanger surfaces. This allows to reduce the risk of local corrosion and corrosion cracking development. A number of VVER plants have conducted full-scale testing that confirmed the impact of ODA microadditions on local corrosion mitigation. Some PWR plants are testing injection of surface-active dispersants to loosen SG deposits. Multiple studies proved ODA ability to remove chlorides from smooth surfaces which allowed to reduce the rate of microcrack growth. Trial testing has shown that the rate of corrosion cracking on SG tubes was reduced by 60-70% owing to ODA injections. Such effect was due to significant reduction in chlorides absorption by the metal surface during the year of ODA injection. Tests on a pilot SG design have shown that ODA could be used for partially wash out deposits from a heating surface. This also minimizes local corrosion. The tests showed that ODA microadditions remove chlorides from microcracks and crevices on SG tubing spacer grids. The ability to wash out previously formed deposits allows to reduce risk of local corrosion and cracking. The abilities of microadditions of film-forming corrosion inhibitors identified through the above mentioned testing could be used under the VVER plant life extension program. The

  12. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  13. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  14. Determination of power density in VVER-1000 Mock-Up in LR-0 reactor

    Science.gov (United States)

    Koš'ál, Michal; Rypar, Vojtìch; Harutyunyan, Davit; Schulc, Martin; Losa, Evžen

    2017-09-01

    The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures.

  15. Analysis of loss of off-site power ATWS in VVER-440 concept

    Energy Technology Data Exchange (ETDEWEB)

    Hoeppner, G.; Siltanen, P.; Kotro, J.

    1987-01-01

    During 1985 the Finnish state-owned utility Imatran Voima Oy signed a work order with Gesellschaft fuer Reaktorsicherheit mbH of the Federal Republic of Germany (GRS) for the analysis of abnormal transients in a pressurized water reactor (PWR) concept based on a Soviet design. The results of these calculations were intended to be introduced into the licensing process and to support a decision to build such a nuclear power station. A computer model was constructed of the VVER-440 concept, a 500-MW(electric) PWR designed in the USSR and modified for Finland. The ALMOD4 code, developed at GRS, was used for the investigation. The ALMOD4 code is a fast running code for the analysis of operational and abnormal transients in PWRs. Input data were set up to calculate anticipated transients without scram, most notably the loss of off-site power case. One-dimensional neutron kinetics was used to correctly model the neutronics feedback of axially distributed moderator density and fuel temperature in a changing axial power profile. Interlocking signals and the engineered safety systems were modeled to assess the overall systems response to this abnormal transient. Special analytical problems were encountered since a detailed and verified model of the steam generator (SG) with horizontally positioned heat exchanger tubes was not available. Therefore, two bounding calculations were performed with different SG models.

  16. A Four Group Reference Code for Solving Neutron Diffusion Equation in a VVER-440 Core

    Energy Technology Data Exchange (ETDEWEB)

    Saarinen, Simo [Fortum Nuclear Services Ltd., P.O. Box 100, 00048 Fortum (Finland)

    2008-07-01

    Nuclear reactor core power calculation is essential in the analysis of the nuclear power plant and especially the core. Currently, the core power distribution in Loviisa VVER-440 core is calculated using nodal code HEXBU-3D and pin-power reconstruction code ELSI-1440 that solve the two group neutron diffusion equation. The computer power available has increased significantly during the last decades allowing us to develop a fine mesh code HEXRE for solving the four group diffusion equation. The diffusion equations are discretized using piecewise linear polynomials. The core is discretized using one node per fuel pin cell. The axial discretization can be chosen freely. The boundary conditions are described using diffusion theory and albedos. Burnup dependence is modelled by tabulating diffusion parameters at certain burnup values and using interpolation for the intermediate values. A two degree polynomial is used for the modelling of the feedback effects. Eigenvalue calculation for both boron concentration and multiplication factor control has been formulated. A possibility to perform fuel loading and shuffling operations is implemented. HEXRE has been thoroughly compared with HEXBU-3D and ELSI-1440. The effect of the different energy and space discretizations used is investigated. Some safety criteria for the core calculated with the HEXRE and HEXBU-3D/ELSI-1440 have been compared. From the calculations (e.g. the safety criteria) we can estimate whether there exists systematic deviations in HEXBU- 3D/ELSI-1440 calculations or not. (author)

  17. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2017-05-15

    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  18. The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine.

    Energy Technology Data Exchange (ETDEWEB)

    Kot, C.

    1999-05-10

    Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Department of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments.

  19. Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

    Science.gov (United States)

    Tyapkov, V. F.; Erpyleva, S. F.

    2017-05-01

    Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6-12 to 2.0-2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains pH25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

  20. Determination of power density in VVER-1000 Mock-Up in LR-0 reactor

    Directory of Open Access Journals (Sweden)

    Košál Michal

    2017-01-01

    Full Text Available The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures.

  1. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50-400)°C

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Bukina, Z. V.; Frolov, A. S.; Maltsev, D. A.; Krikun, E. V.; Zhurko, D. A.; Zhuchkov, G. M.

    2017-07-01

    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50-400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects - dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

  2. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  3. Experimental Testing of Innovative Cold-Formed "GEB" Section / Badania Eksperymentalne Innowacyjnego Kształtownika Giętego Na Zimno Typu „Geb“

    Directory of Open Access Journals (Sweden)

    Łukowicz Agnieszka

    2015-03-01

    Full Text Available Jedną z najważniejszych zalet lekkich konstrukcji metalowych, wytwarzanych z kształtowników giętych na zimno, jest ich mała masa, dlatego też, producenci coraz częściej wykorzystują możliwości profili giętych do wytwarzania typowych konstrukcji halowych w budownictwie systemowym. Proces gięcia na zimno, pozwala na formowanie różnego rodzaju przekrojów poprzecznych, które mogą być wykorzystywane jako elementy konstrukcji. Typowe kształty elementów. tzn. Z, C oraz tzw. przekroje kapeluszowe, które zostały przebadane i opisane w literaturze, wykorzystuje się głównie jako płatwie lub części składowe wiązarów kratowych. Nowo opatentowany przekrój typu GEB ma być wykorzystany jako element nośny konstrukcji ramowych. W związku z tym innowacyjny kształt oraz parametry geometryczne przekroju takiego kształtownika, związane z możliwością jego wyprodukowania oraz z warunkami nośności, stateczności oraz sztywności, muszą być optymalne. Według normy PN-EN 1993-1-3, każdy nowo uformowany przekrój powinien być przebadany pod kątem nośności elementu i formy deformacji, dlatego też, ten innowacyjny kształtownik został poddany m.in. badaniom nośności na zginanie. Na ich podstawie została określona wartość nośności przekroju oraz forma jego deformacji. Jednocześnie dla analizowanego elementu zostały przeprowadzone obliczenia nośności przekroju, według normy PN-EN-1993-1-3 oraz numeryczne w programie MARC MENTAT.

  4. The Design of PSB-VVER Experiments Relevant to Accident Management

    Science.gov (United States)

    Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

  5. IVO participation in IAEA benchmark for VVER-type nuclear power plants seismic analysis and testing

    Energy Technology Data Exchange (ETDEWEB)

    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. 25 refs.

  6. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  7. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-11-01

    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  8. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Directory of Open Access Journals (Sweden)

    Dzhalandinov A.

    2016-01-01

    Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  9. Experimental Testing of Innovative Cold-Formed "GEB" Section / Badania Eksperymentalne Innowacyjnego Kształtownika Giętego Na Zimno Typu "Geb"

    Science.gov (United States)

    Łukowicz, Agnieszka; Urbańska-Galewska, Elżbieta; Gordziej-Zagórowska, Małgorzata

    2015-03-01

    One of the major advantages of light gauge steel structures made of cold-formed steel sections is their low weight so the production of typical single-storey steel structures of this kind of profiles is still rising. The well known profiles, e.o. Z-sections, C-sections and the so called hat-sections studied and described in the literature, are used mainly as purlins or truss components. A new profile GEB was patented for the use for primary load-bearing member in fabricated steel frames. According to the code [1] every novel cross section should be tested to assign the deformation shape and bearing capacity. The paper deals with the numerical and experimental research of bearing capacity of cold formed GEB profiles. The deformation shape and limit load was obtained from bending tests. The GEB cross section bearing capacity was also determined according to codes [1, 2]. Jedną z najważniejszych zalet lekkich konstrukcji metalowych, wytwarzanych z kształtowników giętych na zimno, jest ich mała masa, dlatego też, producenci coraz częściej wykorzystują możliwości profili giętych do wytwarzania typowych konstrukcji halowych w budownictwie systemowym. Proces gięcia na zimno, pozwala na formowanie różnego rodzaju przekrojów poprzecznych, które mogą być wykorzystywane jako elementy konstrukcji. Typowe kształty elementów. tzn. Z, C oraz tzw. przekroje kapeluszowe, które zostały przebadane i opisane w literaturze, wykorzystuje się głównie jako płatwie lub części składowe wiązarów kratowych. Nowo opatentowany przekrój typu GEB ma być wykorzystany jako element nośny konstrukcji ramowych. W związku z tym innowacyjny kształt oraz parametry geometryczne przekroju takiego kształtownika, związane z możliwością jego wyprodukowania oraz z warunkami nośności, stateczności oraz sztywności, muszą być optymalne. Według normy PN-EN 1993-1-3, każdy nowo uformowany przekrój powinien być przebadany pod kątem nośności elementu i formy

  10. Study of non-condensable gases effect on VVER steam generator operation in condensation mode at large-scale facility

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, A. V.; Remizov, O. V.; Tzyganok, A. A.; Kalyakin, D. S., E-mail: sas@ippe.r [Institute for Physics and Power Engineering by A. I. Leypunsky, Bondarenko 1 sq. Obninsk, 249033 Moscow (Russian Federation)

    2010-10-15

    The NNP-2006 project of nuclear power plant with VVER-1200 reactor provides for use of passive safety systems for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (Sg) to operation in the mode of condensation of the primary circuit steam. As a result, the condensate from Sg arrives at the core providing its additional cooling. The joint operation of the PHRS and the system of hydro accumulators of the second stage makes it possible to assure the heat removal from the core during 24 hours. The efficiency of the system can be affected by the presence os non-condensable gases in the primary circuit. The main sources of gases are nitrogen, arriving at the circuit, as hydro accumulators actuate and products of radiolysis of water. The circuit design considered in the project makes it possible to remove gas-steam mixture from Sg. At the same time, it is necessary to ascertain if the gas removal is adequate for ensuring the design operation of Sg in the steam condensation mode. For this purpose, series of experiments have been carried out at the large-scale test facility HA2M-Sg. The test facility incorporates VVER reactor Sg model with volumetric-power scale of piping is 1:46, PHRS heat exchanger imitator and buffer tank, equipped by steam supply system. The elevations of the main equipment correspond to those of reactor project. Experiments at the HA2M-Sg test facility have been performed at the pressure 0.36-0.38 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. The report presents the basic results of experiments aimed at the evaluation of Sg condensation power under the inflow of gas-steam mix to the tube bundle, both under the simulation of gas-steam mixture outflow from Sg and without outflow. (Author)

  11. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  12. Simulation of a control rod ejection in a VVER-1000 reactor with the program Relap5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.; Gehin, J.C.; Yoder, G.L. [Oak Ridge National Lab., TN (United States); Ivanov, V.K. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    2001-07-01

    The RELAP5-3D code has been employed to simulate the ejection of a control rod at the Balakovo-4 plant, a VVER-1000 V320 plant located in Russia. The reactor core contains 163 assemblies, three of them Lead Test Assemblies (LTAs) with mixed oxide (MOX) fuel, and the remaining 160 assemblies with UO{sub 2} fuel. The worth of the ejected control rod was $ 0,225 or 142 pcm. Results from point and three-dimensional (3-D) or nodal kinetics calculations are presented. The results from both models are similar with no significant differences. All calculated results are within safety limits, with no fuel melting or cladding failures predicted to occur. (author)

  13. The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes

    Energy Technology Data Exchange (ETDEWEB)

    Jafari, Naser; Talebi, Saeed [Amirkabir Univ. of Technology, Tehran Polytechnic (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

    2017-07-15

    In this paper, the effect of boron dilution transient, as a consequence of the malfunction of the boron control system, was investigated in a VVER-1000 reactor, and then an appropriate setpoint was determined for the actuation of the emergency protection system to the reactor shutdown. In order to simulate the boron dilution, first, the whole reactor core was simulated by MCNPX code to compute the radial and axial power distribution. Then, the COBRA-EN code was employed using calculated power distribution for analyzing the thermal-hydraulic of hot fuel assembly and for extracting the safety parameters. For the safe operation of the reactor, certain parameters must be in defined specified ranges. Comparison between our results and FSARs data shows that the present modeling provides a good prediction of boron dilution transient with the maximum relative difference about 4%.

  14. Studying the operation of a VVER steam generator in the condensing mode at different parameters of emergency processes

    Science.gov (United States)

    Morozov, A. V.; Shlepkin, A. S.; Kalyakin, D. S.; Soshkina, A. S.

    2017-05-01

    The article presents the results of the experimental study of heat and mass transfer processes in an NPP steam generator during the operation of passive safety systems of new-generation VVER reactor installations. At the GE2M-PG test rig in the Leypunsky Institute for Physics and Power Engineering, two series of experiments corresponding to different stages of the accident were completed. In these experiments, the performance of VVER steam generator in the condensing mode with and without the removal of gas-vapor mixture from the "cold" header has been studied. As a result of the first series of experiments, it was found that, for any of the parameters of the emergency process, the steam generator's power does not drop below 80% of the original value. Furthermore, we revealed that the composition and physical properties of gases in the investigated concentration range did not notably affect the processes in the steam generator. In the second series of experiments without removal of noncondensable gases, the influence of parameters of the emergency process on the efficiency of heat transfer in the steam generator operating in the condensing mode was investigated. In order to study the heat transfer processes, we studied the change of the temperature difference between the media of the first and second circuits in our experiments. We found that the value of the temperature difference depends on both the mass of noncondensable gases accumulated in the tube bundle and their accumulation rate. The accumulation rate is determined by the power of the steam generator and the concentration of gases entering the steam generator. As a result of the analysis of experimental data, we obtained the analytical dependence reflecting change in the power of the steam generator operating in the emergency condensing mode.

  15. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  16. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  17. Comparison of ENDF/B-VII.1 and JEFF-3.2 in VVER-1000 operational data calculation

    Science.gov (United States)

    Frybort, Jan

    2017-09-01

    Safe operation of a nuclear reactor requires an extensive calculational support. Operational data are determined by full-core calculations during the design phase of a fuel loading. Loading pattern and design of fuel assemblies are adjusted to meet safety requirements and optimize reactor operation. Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors. Nuclear data for this diffusion code are prepared regularly by lattice code HELIOS. These calculations are conducted in 2D on fuel assembly level. There is also possibility to calculate these macroscopic data by Monte-Carlo Serpent code. It can make use of alternative evaluated libraries. All calculations are affected by inherent uncertainties in nuclear data. It is useful to see results of full-core calculations based on two sets of diffusion data obtained by Serpent code calculations with ENDF/B-VII.1 and JEFF-3.2 nuclear data including also decay data library and fission yields data. The comparison is based directly on fuel assembly level macroscopic data and resulting operational data. This study illustrates effect of evaluated nuclear data library on full-core calculations of a large PWR reactor core. The level of difference which results exclusively from nuclear data selection can help to understand the level of inherent uncertainties of such full-core calculations.

  18. Comparison of ENDF/B-VII.1 and JEFF-3.2 in VVER-1000 operational data calculation

    Directory of Open Access Journals (Sweden)

    Frybort Jan

    2017-01-01

    Full Text Available Safe operation of a nuclear reactor requires an extensive calculational support. Operational data are determined by full-core calculations during the design phase of a fuel loading. Loading pattern and design of fuel assemblies are adjusted to meet safety requirements and optimize reactor operation. Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors. Nuclear data for this diffusion code are prepared regularly by lattice code HELIOS. These calculations are conducted in 2D on fuel assembly level. There is also possibility to calculate these macroscopic data by Monte-Carlo Serpent code. It can make use of alternative evaluated libraries. All calculations are affected by inherent uncertainties in nuclear data. It is useful to see results of full-core calculations based on two sets of diffusion data obtained by Serpent code calculations with ENDF/B-VII.1 and JEFF-3.2 nuclear data including also decay data library and fission yields data. The comparison is based directly on fuel assembly level macroscopic data and resulting operational data. This study illustrates effect of evaluated nuclear data library on full-core calculations of a large PWR reactor core. The level of difference which results exclusively from nuclear data selection can help to understand the level of inherent uncertainties of such full-core calculations.

  19. Almost twenty years' search of transuranium isotopes in effluents discharged to air from nuclear power plants with VVER reactors.

    Science.gov (United States)

    Hölgye, Z; Filgas, R

    2006-04-01

    Airborne effluents of 5 stacks (stacks 1-5) of three nuclear power plants, with 9 pressurized water reactors VVER of 4,520 MWe total power, were searched for transuranium isotopes in different time periods. The search started in 1985. The subject of this work is a presentation of discharge data for the period of 1998-2003 and a final evaluation. It was found that 238Pu, 239,240Pu, 241Am, 242Cm, and 244Cm can be present in airborne effluents. Transuranium isotope contents in most of the quarterly effluent samples from stacks 2, 4 and 5 were not measurable. Transuranium isotopes were present in the effluents from stack l during all 9 years of the study and from stack 3 since the 3rd quarter of 1996 as a result of a defect in the fuel cladding. A relatively high increase of transuranium isotopes in effluents from stack 3 occurred in the 3rd quarter of 1999, and a smaller increase occurred in the 3rd quarter of 2003. In each instance 242Cm prevailed in the transuranium isotope mixtures. 238Pu/239,240Pu, 241Am/239,240Pu, 242Cm/239,240Pu, and 244Cm/239,240Pu ratios in fuel for different burn-up were calculated, and comparison of these ratios in fuel and effluents was performed.

  20. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  1. Experimental investigation of non-equilibrium thermal-hydraulic processes in a new passive VVER core reflooding system

    Energy Technology Data Exchange (ETDEWEB)

    Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov; Alexandr A Tzyganok [State Scientific Center of Russian Federation - Institute for Physics and Power Engineering by A.I. Leypunsky, IPPE, 1, Bondarenko sq., Obninsk, 249030 (Russian Federation)

    2005-07-01

    Full text of publication follows: In the systems of passive reactor cooling and core reflooding from above under accident conditions, the cold water flows out into countercurrent steam flow; that is, the direct contact between steam and liquid occurs simultaneously over all cross-sections of pipes/closed spaces. The process is complicated by additional effect of steam condensation and droplet flow towards the steam flow. The processes of heat and mass transfer proceed simultaneously at variable liquid level. These factors give rise to a nonsteady, non-equilibrium two-phase flow. The operating conditions of such a non-steady process, the main of which being the time of closed water volume dump (or the flow coefficient), can be obtained only experimentally. The present work is devoted to the experimental investigations of the interaction of saturated steam with cold water at its flowing out from a closed space with a variable level and some characteristics of dynamic two-phase layer at the steam/liquid interface. With reference to the system of passive heat removal from VVER core, the processes of interaction of saturated steam, steam-water mixture, and air with cold water at its flowing out from a vertical plugged pipe with an internal diameter of 50 and 100 mm have been studied at a pressure of 0.5 MPa. It has been stated experimentally that the dump rate of subcooled water from a plugged pipe into steam is nearly an order of magnitude less than that into non-condensable gas media. The semi-empirical correlation describing the processes of water outflow from plugged pipes into steam is of the form of: W-bar = C{sub 0} {radical}(gd), where C{sub 0} is the dimensionless constant, d is the pipe internal diameter, m, g = 9.81 m/s{sup 2}. (authors)

  2. Duckweed does not improve the efficiency of municipal wastewater treatment in lemna system plants / Rzęsa nie poprawia efektywności oczyszczania ścieków komunalnych w oczyszczalniach typu Lemna System

    Directory of Open Access Journals (Sweden)

    Ozimek Teresa

    2015-09-01

    Full Text Available Badania prowadzono w trzech czyszczalniach hydrofi towych z powierzchniowym przepływem ścieków typu Lemna System, zbudowanych według projektu Lemna Corporation. Składają się one ze stawu napowietrzanego i stawu rzęsowego wyposażonego w system barier mających na celu równomierne rozmieszczenie roślin. Stawy poprzedzone są osadnikiem wstępnym. Według projektantów odpowiednia praca tych oczyszczalni związana jest z występowaniem rzęsy na powierzchni stawu rzęsowego. Badano efektywność oczyszczania ścieków komunalnych w trzech oczyszczalniach Lemna System usytuowanych w centralnej Polsce ze szczególnym uwzględnieniem roli stawu rzęsowego i samej rzęsy. Oczyszczalnie różniły się między sobą występowaniem rzęsy. W dwóch z nich (w Rakowie i Bąkowcu w trakcie całego okresu badań rzęsa nie występowała, natomiast w trzeciej (w Falęcinie Starym w okresie wegetacyjnym pokrywała do 90%.powierzchni stawu. Efektywność oczyszczanie ścieków nie różniła się w oczyszczalni z rzęsą i bez rzęsy. Rzęsa nie wpływała istotnie na efektywność oczyszczania ścieków. We wszystkich oczyszczalniach główną rolę w redukcji zanieczyszczeń odgrywał staw napowietrzany

  3. Influence of operation factors on brittle fracture initiation and critical local normal stress in SE(B) type specimens of VVER reactor pressure vessel steels

    Science.gov (United States)

    Kuleshova, E. A.; Erak, A. D.; Kiselev, A. S.; Bubyakin, S. A.; Bandura, A. P.

    2015-12-01

    A complex of mechanical tests and fractographic studies of VVER-1000 RPV SE(B) type surveillance specimens was carried out: the brittle fracture origins were revealed (non-metallic inclusions and structural boundaries) and the correlation between fracture toughness parameters (CTOD) and fracture surface parameters (CID) was established. A computational and experimental method of the critical local normal stress determination for different origin types was developed. The values of the critical local normal stress for the structural boundary origin type both for base and weld metal after thermal exposure and neutron irradiation are lower than that for initial state due to the lower cohesive strength of grain boundaries as a result of phosphorus segregation.

  4. The influence of operational and water chemistry parameters on the deposits of corrosion products on fuel assemblies at nuclear power plants with VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Rodionov, Yu. A.; Gavrilov, A. V.

    2011-07-01

    The phenomenon involving a growth of pressure drop in the reactor core and redistribution of deposits in the reactor core and primary coolant circuit of a nuclear power station equipped with VVER-440 reactors is considered. A model is developed, the physicochemical foundation of which is based on the dependence of corrosion product transfer on the temperature and pH t value of coolant and on the correlation between the formation rate of corrosion products (Fe) (after subjecting the steam generators to decontamination) and rate with which they are removed from the circuit. The purpose of the simulation carried on the model is to predict the growth of pressure drop on the basis of field data obtained from nuclear power installations and correct the water chemistry (by adjusting the concentrations of KOH, H2, and NH3) so as to keep the pressure drop in the reactor at a stable level.

  5. Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany); Nikonov, S. [NRC ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2014-08-15

    The paper presents an uncertainty and sensitivity (U and S) study of the VVER-1000 reactor hydraulic properties. It is based on the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). The novelty of the work consists of taking into consideration all hydraulic uncertainty parameters used in the modeling of the reactor pressure vessel (RPV) internals. A detailed parallel channel ATHLET model of the RPV is developed. It consists of ca. 26 600 control volumes most of them connected with junctions for cross flows. The specific geometry of the gap between upper part of the baffle and upper part of fuel assembly and also a fuel assembly head is taken explicitly into account The influence of the input parameters on the sensitivity and uncertainty of the RPV outlet and inlet temperatures and mass flows as well assembly-wise mass flow and coolant temperature axial distributions is shown.

  6. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool; L'essai Plinius/Colima CA-U3 sur le relachement des aerosols de produits de fission au-dessus d'un bain de corium de type VVER

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L

    2007-07-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  7. Piec mitow o badaniach typu studium przypadku

    DEFF Research Database (Denmark)

    Flyvbjerg, Bent

    2005-01-01

    useful for generating hypotheses, while other methods aremore suitable for hypotheses testing and theory building; (4) The case study contains a bias toward verification; and (5) It is often difficult to summarize specific case studies. The article explains and corrects these misunderstandings one by one...

  8. Start-up of a cold loop in a VVER-440, the 7{sup th} AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

    Energy Technology Data Exchange (ETDEWEB)

    Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

    2017-09-15

    The 7{sup th} dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7{sup th} AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7{sup th} AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

  9. Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)

    2007-09-15

    Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.

  10. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Energy Technology Data Exchange (ETDEWEB)

    Rahmani, Yashar, E-mail: yashar.rahmani@gmail.com [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)

    2017-03-15

    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  11. The passive system for reflooding of the VVER reactor core from the second-stage hydro-accumulators: design and basic design solutions

    Energy Technology Data Exchange (ETDEWEB)

    Alexandr D Efanov; Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov [State Scientific Center of Russian Federation - Institute for Physics and Power Engineering by A.I. Leypunsky, IPPE, 1, Bondarenko sq., Obninsk, 249030 (Russian Federation); Vladimir M Berkovich; Victor N Krushelnitskiy; Vladimir G Peresadko [FSUE Atomenergoproekt, B-5, 1 bldg.7, Bakuninskaya, 107005, Moscow (Russian Federation); Yuri G Dragunov; Alexey K Podshibyakin; Sergey I Zaitcev [FSUE OKB Gidropress, 21, Ordzhonikidze street, 142103 Podolsk (Russian Federation)

    2005-07-01

    Full text of publication follows: The fundamental difference in the safety assurance of the operating NPPs and those under design implies that the safety in the existing NPPs is achieved by energy-dependent (active) systems and depends on the proficiency of attending personnel. To provide safety, the new NPP designs use the physical processes proceeding in the facility without power supply; and they are unaffected by human errors. As to the safety level, the design of the new generation nuclear power plant NPP-92 relates to the class of the improved NPPs; and it applies a principle of diversity in the structure of systems responsible for critical safety functions. In accordance with the above-mentioned safety concept, the design development required a complex of experimental investigations and numerical modeling to be conducted. Among the passive safety systems of the NPP with RP-392 is the system of the second stage hydro-accumulators (GE-2). The system of the second-stage hydro-accumulators consists of four groups of hydro-accumulating tanks with a total coolant volume of 960 m{sup 3}. The system is intended for the core flooding with coolant during 24 hours. In each group of the hydro-accumulators, the graded coolant flowrate is provided, which depends on residual heat in the reactor. The special check valves are tuned to open at the pressure drop in the circuit below 1.5 MPa. The paper presents the thermalhydraulic substantiation of the serviceability of the second-stage hydro-accumulators system for passive heat removal from the VVER reactor core and the basic design solutions on the GE-2 system. (authors)

  12. Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

    Science.gov (United States)

    Parchevsky, V. M.; Guryanova, V. V.

    2017-01-01

    A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGB that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the "hot" header on the water level the "cold" end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.

  13. Types and analysis of defects in welding junctions of the header to steam generator shells on power-generating units with VVER-1000

    Science.gov (United States)

    Ozhigov, L. S.; Voevodin, V. N.; Mitrofanov, A. S.; Vasilenko, R. L.

    2016-10-01

    Investigation objects were metal templates, which were cut during the repair of welding junction no. 111 (header to the steam generator shell) on a power-generating unit with VVER-1000 of the South-Ukraine NPP, and substances of mud depositions collected from walls of this junction. Investigations were carried out using metallography, optical microscopy, and scanning electron microscopy with energy dispersion microanalysis by an MMO-1600-AT metallurgical microscope and a JEOL JSM-7001F scanning electron microscope with the Shottky cathode. As a result of investigations in corrosion pits and mud depositions in the area of welding junction no. 111, iron and copper-enriched particles were revealed. It is shown that, when contacting with the steel header surface, these particles can form microgalvanic cells causing reactions of iron dissolution and the pit corrosion of metal. Nearby corrosion pits in metal are microcracks, which can be effect of the stress state of metal under corrosion pits along with revealed effects of twinning. The hypothesis is expressed that pitting corrosion of metal occurred during the first operation period of the power-generating unit in the ammonia water chemistry conditions (WCC). The formation of corrosion pits and nucleating cracks from them was stopped with the further operation under morpholine WCC. The absence of macrocracks in metal of templates verifies that, during operation, welding junction no. 111 operated under load conditions not exceeding the permissible ones by design requirements. The durability of the welding junction of the header to the steam generator shell significantly depends on the technological schedule of chemical cleaning and steam generator shut-down cooling.

  14. Behavior of a VVER fuel element tested under severe accident conditions in the CORA facility. Test results of experiment CORA-W1

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, S.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.; Sepold, L.

    1994-01-01

    Test bundle CORA-W1 was without absorber material. As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the test were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zirconium/niobium-steam reaction started at about 1200 C, leading the bundle to a maximum temperature of approximately 1900 C. With the movement of the melt also heat is transported to the lower region. Below 300 mm elevation the test bundle remained intact due to the axial temeprature distribution. W2 ist characterized by a strong oxidation above 300 mm elevation. Besides the severe oxidation the test bundle resulted in considerable fuel dissolution by ZrNb1/UO{sub 2} interaction in the upper part, complete spacer destruction at 600 mm due to chemical interactions between steel and the ZSrNb1 cladding. Despite some specific features the material behavior of the VVER-1000 bundle is comparable to that observed in the PWR and BWR test using fuel elements typical for Western countries. (orig./HP) [Deutsch] Versuchsbuendel CORA-W1 hatte kein Absorberelement. Wie in den CORA-Versuchen zuvor wurden die Testbuendel in Dampfatmosphaere Temperaturtransienten mit langsamer Aufheizrate ausgesetzt. Damit wurde ein Unfallablauf fuer einen LWR simuliert, der sich aus einem Kuehlmittelverluststoerfall durch Auftreten eines sogenannten kleinen Lecks entwickeln kann. Die Temperatureskalation - aufgrund der exothermen Zirkon/Niob-Wasserdampfreaktion - setzte ab ca. 1100 C ein. Die Hoechsttemperaturen im Buendel betrugen 2000 C. Die Versuchsdaten des Experiments CORA-W1 werden zusammen mit ersten Ergebnissen der Nachuntersuchung dargelegt. Das Versuchsbuendel weist eine starke Oxidation im oberen Bereich auf. Der untere Buendelabschnitt (bis 400 mm) blieb aufgrund der niedrigen Temperaturen in diesem

  15. Inter-comparison of JEF-2.2 and JEFF-3.1 evaluated nuclear data through Monte Carlo analysis of VVER-1000 MOX Core Computational Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Thilagam, L., E-mail: thilagam@igcar.gov.i [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India); Karthikeyan, R., E-mail: rkarthi@barc.gov.i [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Jagannathan, V., E-mail: v_jagan1952@rediffmail.co [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Subbaiah, K.V.; Lee, S.M. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)

    2010-02-15

    The nuclear data forms a vital component in reactor core physics computations. The nuclear data is evaluated and modified on a continuous basis by different nuclear data centres and laboratories worldwide. The work on upgradation of the nuclear data is being carried out using new evaluations obtained through experiments and theoretical models to enhance their accuracy. Use of different sets of cross-section data in the analysis of a benchmark problem is a source of strong feedback for further improvements in data by mutual comparison of results. These comparisons also help to find out the best evaluated cross-section data released. Towards this objective, an attempt has been made to inter-compare JEF-2.2 and JEFF-3.1 evaluated nuclear data through the Monte Carlo simulation of 'VVER-1000 MOX Core Computational Benchmark'. This study deals with the calculation and inter-comparison of reactor parameters such as multiplication factors, cell average and assembly average fission reaction rate distributions estimated for various reactor state descriptions specified in the benchmark. Point-wise cross-section libraries processed from the JEF-2.2 and JEFF-3.1 evaluated data are used in the analysis. Concerning the multiplication factors and fission rate distributions, considerable differences are observed between the two libraries. While performing the MCNP calculations with JEFF-3.1 data, it is observed that the deviations of effective neutron multiplication factors (k{sub eff}) from those of benchmark standard MCU results are lower by about approx0.100% for the most of the states than those computed using JEF-2.2. Fission rate distributions using JEFF-3.1 data are also found to have significant deviations up to +-9.2% compared to calculations with its earlier version JEF-2.2 data. Some interesting trends on the used nuclear data are identified from the discrepancies of the individual results. The cause for considerable changes in the calculated parameters are

  16. Występowanie cech kofeinizmu u młodzieży i młodych dorosłych deklarujących częste spożycie bezalkoholowych napojów typu cola zawierających kofeinę

    Directory of Open Access Journals (Sweden)

    Sandra Gustek

    2011-12-01

    Full Text Available Wprowadzenie: W ciągu ostatnich lat obserwuje się znaczny wzrost spożycia napojów zawierających kofeinę, szczególnie wśród dzieci i młodzieży. Tendencja ta jest niepokojąca, a liczne badania wskazują na negatywny wpływ nadmiernej podaży kofeiny na różne aspekty funkcjonowania człowieka. W celu głębszej analizy tego zjawiska przeprowadzono badanie własne i dokonano próby oceny problemu uzależnienia od kofeiny wśród młodzieży i młodych dorosłych ze szczególnym uwzględnieniem cech kofeinizmu u osób deklarujących czę- ste spożycie napojów typu cola. Materiał i metodyka: Na podstawie przeglądu literatury dotyczącej kofeinizmu skonstruowano autorski kwestionariusz ankiety. Określono następujące kryteria włączenia do badania: 1 deklaracja o spożywaniu napojów typu coca-cola, 2 wiek 13-30 lat. W badaniu metodą sondażu diagnostycznego wzięło udział 118 osób. Dodatkowo kontrolowano także zmienne demograficzne, tj. wiek, płeć oraz wzrost i masę ciała. Wyniki: Najczęściej deklarowane przez respondentów potencjalne objawy kofeinizmu to: bóle głowy, drażliwość i wybuchowość, problemy ze snem, przyspieszona praca serca i palpitacje, przy czym symptomy te uzależnione były od wieku. Wnioski: Nadmierne spożycie produktów zawierających kofeinę może być ważnym czynnikiem środowiskowym w kształtowaniu i utrzymywaniu się chorób przewlekłych oraz uzależnień, a także wiązać się ze złymi zachowaniami żywieniowymi. Dlatego tak istotne są profilaktyka i zwiększanie świadomości dzieci, młodzieży oraz ich rodziców na temat szkodliwości nadmiernej podaży produktów zawierających kofeinę, a także dalsze badania nad wpływem kofeiny na funkcjonowanie psychofizyczne tej grupy wiekowej.

  17. Validation of computer codes and modelling methods for giving proof of nuclear safety of transport and storage of spent VVER-type nuclear fuels. Pt. 2. Criticality safety during transport and storage of spent VVER fuel elements. Final report; Einschaetzung von Rechenprogrammen und Methoden zum Nachweis der nuklearen Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoffen. T. 2. Kritikalitaetssicherheit bei Transport und Lagerung von WWER-Brennelementen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-03-15

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k{sub eff} is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [Deutsch] Der Bericht enthaelt die Ergebnisse von Untersuchungen zur Validierung von Rechenprogrammen, welche zum Nachweis der nukelaren Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoff der KKW Greifswald und Rheinsberg eingesetzt wurden. Es werden eine Charakteristik des abgebrannten Brennstoffs (Nuklidkonzentrationen, Neutronenquellstaerke, Gammaspektrum, Nachzerfallsleistung) - berechnet mit verschiedenen Programmen - und Ortsdosisleistungen (z.B. in der Umgebung eines Transportbehaelters) - basierend auf den verschiedenen Quelltermen - angegeben. Differenzen und Ursachen werden diskutiert. Die Ergebnisse zeigen, dass trotz der Differenzen in den Quelltermen alle strahlenschutztechnisch relevanten Aussagen unbeeinflusst bleiben. Fuer die Einschaetzung des Gueltigkeitsbereiches des Monte-Carlo-Programms OMEGA wurden ca. 200 kritische Experimente mit LWR-Brennstoff unter besonderer Beruecksichtigung

  18. STC Germany/Russia. Fluence calculations of surveillance specimens of the VVER-440. Final report; WTZ Russland. Fluenzberechnungen fuer Voreilproben beim WWER-440. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Konheiser, J.; Grahn, A.

    2014-07-01

    Reactor pressure vessels (RPV) are non-restorable equipment and their lifetime may restrict the nuclear power plant-life as a whole. Surveillance specimen programs for RPV materials are among the most important measures of in-service inspection programs that are necessary for realistic and reliable assessment of the RPV residual lifetime. In addition to the chemical composition of the RPV steel, the radiation parameters (neutron and gamma fluences and spectra) have the most important impact on the RPV embrittlement characteristics. In this work, different geometric positions which have influence on the radiation conditions of the samples are investigated. Thus, the uncertainties can be determined in the fluence values of surveillance specimens. The fluence calculations were carried out by the codes TRAMO and DORT. This study was accompanied by ex-vessel neutron dosimetry experiments at Kola NPP, Unit 3 (VVER-440/213), which provide the basis for validation of calculated neutron fluences. The main neutron-activation monitoring reactions were {sup 54}Fe(n,p){sup 54}Mn and {sup 58}Ni(n,p){sup 58}Co. The activity measurements were carried out by ''Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS). Good agreement between the deterministic and stochastic calculation results as well as between the calculations and the ex-vessel measurements was found. The average difference between measured and calculated values is 5%. The influence of the channels for surveillance specimens and the shielding effect of a baffle rib on the monitors and on the Monte-Carlo calculated results was studied. For the surveillance specimens in the maximum of the flux, an average flux of around 2.45 * 10{sup 12} neutrons/cm{sup 2} was calculated for the neutron flux E> 0.5 MeV. The differences in the surveillance specimens could be up to 20% depending on the direction to the core. Discrepancies up to 10% can be caused by the change of the position of the

  19. Arablwalencja rodzaju rzeczowników typu kieszeń

    OpenAIRE

    Książek-Bryłowa, Władysława

    1981-01-01

    The present study is based on Polish vocabularies and dictionaries and dialect data. It describes gender shifts in some nouns, particularly in consonantal, feminine non-adjectival ones. The conclusion is that the changes in those nouns are of two kinds; 1) they are transferred to masculine group of N nouns (and, at the same time, in some masculine nouns they tend to change the gender into feminine), and 2) they take on the ending -a, typical of feminine nouns.. The pap...

  20. Microstructure alterations in the base material, heat affected zone and weld metal of a 440-VVER-reactor pressure vessel caused by high fluence irradiation during long term operation: material: 15 Ch2MFA {approx} 0, 15 C-2,5 Cr-0, 7Mo-0,3 V; Veraenderungen der Mikrostruktur in Grundwerkstoff, WEZ und Schweissgut eines 440-VVER-Reaktordruckbehaelters, verursacht durch Neutronenbestrahlung im langzeitigen Betrieb; Werkstoff: 15 Ch2MFA {approx} 0,15 C-2,5 Cr-0, 7Mo-0,3 V

    Energy Technology Data Exchange (ETDEWEB)

    Maussner, G.; Scharf, L.; Langer, R. [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Gurovich, B. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1998-11-01

    Within the scope of the Tacis `91/1.1 project of the European Community, ``Reactor Vessel Embrittlement``, specimens were taken from the heavily irradiated circumferential welds of a VVER pressure vessel. The cumulated fast neutron fluence in the specimens amounts to up to 6.5 x 10{sup 19} cm{sup -}2 (E > 0.5 MeV). For the multi-laboratory, coordinated study, the specimens were cutted for mechanical testing as well as analytical, microstructural and microanalytical examinations in the base metal, HAZ and weld metal with respect to the effects of reactor operatio and post-irradiation annealing as well as thermal treatment (475 C, 560 C). The analytical transmission electron microscopy (200 kV) revealed the alterations found in the mechanical properties to be due to the formation of black dots and irradiation-induced segregations and accumulations of copper and carbides. These effects, caused by operation, (neutron radiation, temperature), are much more significant in the HAZ than in the base metal. (orig./CB) [Deutsch] Im Rahmen des von der Europaeischen Union beauftragten Tacis `91/1.1 Programms `Reactor Vessel Embrittlement` wurden Bohrkerne aus dem hochbestrahlten Rundnahtbereich eines VVER-Reaktordruckbehaelters entnommen. Die kumulierte schnelle Neutronenfluenz in diesen Proben betraegt bis zu 6,5 x 10{sup 19} cm{sup -2} (E>0,5 MeV). In einer gemeinschaftlichen Untersuchung wurden mechanisch-technologische, chemische sowie mirkostrukturelle Untersuchungen an Grundwerkstoff-, WEZ- und Schweissgutproben im vergleichbaren Ausgangs-, bestrahlten und anschliessend waermebehandelten (475 C, 560 C) Werkstoffzustand durchgefuehrt. Die analytische Durchstrahlelektronenmikroskopie (200 kV) laesst als Ursache fuer die festgestellten Veraenderungen der mechanischen Eigenschaften die Bildung von Versetzungsringen (black dots) sowie von bestrahlungsinduzierten Ausscheidungen und Anreicherungen von Kupfer in den Karbiden erkennen. Diese Effekte, als Folge der betrieblichen

  1. Pressure Vessel Calculations for VVER-440 Reactors

    Science.gov (United States)

    Hordósy, G.; Hegyi, Gy.; Keresztúri, A.; Maráczy, Cs.; Temesvári, E.; Vértes, P.; Zsolnay, É.

    2003-06-01

    Monte Carlo calculations were performed for a selected cycle of the Paks NPP Unit II to test a computational model. In the model the source term was calculated by the core design code KARATE and the neutron transport calculations were performed by the MCNP. Different forms of the source specification were examined. The calculated results were compared with measurements and in most cases fairly good agreement was found.

  2. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  3. Pracovní využití teriérů typu bull

    OpenAIRE

    TÖRÖKOVÁ, Jacquelina

    2013-01-01

    This work deals with the history of born, origin, domestication of the dog and various types of bull terriers. Pointing at the crossing of the Bulldog with a black terrier. This work also deals with the different types of dogs Bull, their character, description. This work answers the question, ?Why the pit bull is not dangerous to people", evaluates conflict situations, and shows how conflicts can occur. It mentions the current differences and cons of each breed. It places an attempt to creat...

  4. Welding of Prosthetic Alloys / Spawanie Stopów Protetycznych Typu Co-Cr

    Directory of Open Access Journals (Sweden)

    Wojciechowska M.

    2015-12-01

    Full Text Available This paper presents the techniques of joining metal denture elements, used in prosthetic dentistry: the traditional soldering technique with a gas burner and a new technique of welding with a laser beam; the aim of the study was to make a comparative assessment of the quality of the joints in view of the possibility of applying them in prosthetic structures. Fractographic examinations were conducted along with tensile strength and impact strength tests, and the quality of the joints was assessed compared to the solid metal. The experiments have shown that the metal elements used to make dentures, joined by the technique which employs a laser beam, have better strength properties than those achieved with a gas burner.

  5. Tvorba aplikací typu klient/server pomocí Windows Communication Foundation

    OpenAIRE

    KAFKA, Petr

    2010-01-01

    This bachelor work deals with usage of Windows Communication Foundation technology to create application of client/server type. The main goal is to create learning materials, which will familiarize reader with creating WCF clients and services, written in Czech language. These materials contain a number of samples used to explain the problems.

  6. Otěr minerálních katalyzátorů ve fluidním zplyňovacím reaktoru

    Czech Academy of Sciences Publication Activity Database

    Hartman, Miloslav; Svoboda, Karel; Pohořelý, Michael; Šyc, Michal

    2012-01-01

    Roč. 106, č. 9 (2012), s. 844-846 ISSN 0009-2770 R&D Projects: GA AV ČR IAA400720701 Grant - others:RFCR(XE) CT2010-00009 Institutional support: RVO:67985858 Keywords : gasification of biomass * catalyst attrition * dolomites and limestones Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 0.453, year: 2012

  7. Spalování kukuřičné slámy v reaktoru s bublinovou fluidní vrstvou.

    OpenAIRE

    Durda, T. (Tomáš); Moško, J. (Jaroslav); Pohořelý, M; Svoboda, K; Zach, B. (Boleslav); Šyc, M. (Michal); Jeremiáš, M. (Michal); Krausová, A. (Aneta); Punčochář, M.

    2016-01-01

    The paper deals with low-temperature bubbling fluidized-bed combustion of corn straw with very low melting point of ash. The research study was focused on influence of change of selected parameters on emissions of pollutants. Parameters that were changed within the tests were: primary fluidized bed material, combustion medium composition and concentration of oxygen in combustion medium. In order to observe influence of investigated combustion characteristics on emissions of pollutants and on ...

  8. CFD analysis and flow model reduction for surfactant production in helix reactor = CFD analiza i redukcija modela strujanja za proizvodnju surfaktanta u helix reaktoru

    NARCIS (Netherlands)

    Nikačević, N.M.; Thielen, L.; Twerda, A.; Hof, P.M.J. van den

    2015-01-01

    Flow pattern analysis in a spiral Helix reactor is conducted, for the application in commercial surfactant production. Step change response curves (SCR) were obtained from numerical tracer experiments by three-dimensional computational fluid dynamics (CFD) simulations. Non-reactive flow is

  9. Department of Energy's team's analyses of Soviet designed VVERs

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document provides Appendices A thru K of this report. The topics discussed respectively are: radiation induced embrittlement and annealing of reactor pressure vessel steels; loss of coolant accident blowdown analyses; LOCA blowdown response analyses; non-seismic structural response analyses; seismic analyses; S'' seal integrity; reactor transient analyses; fire protection; aircraft impacts; and boric acid induced corrosion. (FI).

  10. SYSTEM WYŚWIETLANIA PARAMETRÓW LOTU DLA ŚMIGŁOWCA TYPU Mi-17 – WYNIKI BADAŃ

    Directory of Open Access Journals (Sweden)

    Bęczkowski Grzegorz

    2015-08-01

    Full Text Available Przedmiotem artykułu są badania w locie nowo opracowanego systemu wyświetlania parametrów lotu (SWPL. Omawiany system przeznaczony jest do zobrazowania na wyświetlaczu przeziernym umieszczonym przed okiem pilota pełnej informacji pilotażowo-nawigacyjnej, ostrzeganiu o sytuacji niebezpiecznej na pokładzie śmigłowca i sygnalizacji błędów pracy systemów pokładowych. System przystosowany jest do pracy w warunkach dziennych i nocnych, przy czym w warunkach nocnych współpracuje z goglami noktowizyjnymi stosowanymi w lotnictwie Sił Zbrojnych RP.

  11. Laboratorní úloha Virtuální sítě typu OpenVPN

    OpenAIRE

    Kortus, Jiří

    2012-01-01

    Tato práce v úvodu popisuje principy virtuálních privátních sítí (VPN) a stručně zmiňuje hlavní protokoly a přístupy pro jejich realizaci. Následně se podrobněji zaměřuje na seznámení čtenáře s charakterem a vlastnostmi sítí OpenVPN. Uvádí jednotlivé rysy a možnosti, diskutuje jejich výhody a nevýhody. Dále se věnuje návrhu laboratorní práce zaměřené na použití OpenVPN. Diskutuje možnosti návrhu práce zejména z technického a didaktického hlediska. Na jejich základě představuje koncept laborat...

  12. O wrażliwośco kontekstowej zdań typu 'S wie, że p' (ON CONTEXTUAL SENSITIVITY OF SENTENCE TYPE 'S KNOWS THAT P'

    Directory of Open Access Journals (Sweden)

    Rafał Palczewski

    2005-12-01

    Full Text Available Semantic contextualism is recently one of the most discussed epistemological theories. So far, the main part of discussion concerns contextualist solution of the skeptical problem. Nowadays it has become more clear that this theory needs strong and independent justification from a linguistic and language-philosophical point of view. In this paper The author outlines several treads concerning linguistic basis for contextualism. In part one there are presented some fundamental contextualism thesis and an example proposed by S. Cohen which has to support it. Next the following question is considered: which semantic feature is responsible for context dependence of knowledge ascription sentences? Is it indexicality, vagueness, ambiguity, ellipticity or unspecificity? Debate sketched in part three is concentrated on an analogy between linguistic behavior of knowledge ascription sentences and other context-dependent expressions, especially indexicals and gradable adjectives. The last part of this paper contains a new argument for contextualism proposed by K. De Rose. In addition such argument points out that contextualism does not confuse a truth conditions of knowledge ascribing sentences with their condition of warranted assertability.

  13. Ecological and economic aspects of programmers uses for regulation of boilers type WR-25; Ekologiczno-ekonomiczne aspekty zastosowania sterownikow mikroprocesorowych do regulacji kotlow typu WR-25

    Energy Technology Data Exchange (ETDEWEB)

    Popiolkiewicz, R. [Zaklad Badawczo-Doswiadczalny Gospodarki Komunalnej, Katowice (Poland); Reron, Z. [Przedsiebiorstwo Energetyki Cieplnej, Dabrowa Gornicza (Poland)

    1995-07-01

    In this article a process of combustion in the boiler is described. It was used automatic control of blast air. A specification of boiler WR-25 is given. Automatic control of blast air gives rational economic effects for example: reduction of combustion fuel and this cause limitation of gases and dust emission. 2 refs., 3 figs., 4 tabs.

  14. Výstavba a provoz sportovního areálu pro jízdu v terénu na MTB typu single trek.

    OpenAIRE

    Novák, Jan

    2012-01-01

    Bakalářská práce se věnuje projektu výstavby sportovního areálu jízdy na horských kolech v terénu pro širokou veřejnost. První část práce se zabývá obecným rozdělením pohybových aktivit a jejím zdravotním přínosem. Dále se práce zabývá významem horské cyklistiky a novým trendem MTB cyklistiky single trek. V dalším kroku práce rozebírá podnikatelský záměr výstavby single trek a přínosem pro danou lokalitu. This bachelor's thesis deals project construction of mountain bikes facilities in ter...

  15. Wpływ polimorfizmu wirusa zapalenia wątroby typu B na przebieg choroby u osób przewlekle zakażonych

    Directory of Open Access Journals (Sweden)

    Magda Rybicka

    2011-01-01

    Full Text Available Hepatitis B virus (HBV infection is one of the major human health problems worldwide. It is estimated that chronic HBV infection affects more than 350 million people globally. It is one of the leading causes of liver cirrhosis and hepatocellular carcinoma. High genetic variability is a characteristic feature of HBV as the viral polymerase lacks proofreading activity. The nucleotide substitution rate for HBV is 10-fold higher than for other DNA viruses.Genetic variations of HBV influence the clinical outcome of HBV infection. There are eight genotypes of hepatitis B virus (A–H that have a distinct geographical distribution. There is clinical significance of HBV genotype in terms of disease activity, risk of progression to cirrhosis, the development of hepatocellular carcinoma and response to antiviral treatments. Moreover, polymorphism in HBV viral polymerase influences the development of HBV mutants resistant to nucleotide analogue treatment that is a consequence of treatment failure.

  16. Porovnání vín školených v různých sudech typu barrique

    OpenAIRE

    Čevela, Jakub

    2016-01-01

    The diplom thesis deals with comparation od the wines aged in different types of barrique barrels. In the theoretical part there are described types and origin of the wood, processing of the wood and chemical compounds, which are brought into the wine during aeging. Practical part deals with comparation of the data from spectrophotometry and chromatography analysis.

  17. Životní styl seniorů s diabetem mellitem 2. typu s ohledem na výživu a pohybovou aktivitu

    OpenAIRE

    Petrová, Zuzana

    2017-01-01

    Recently, the number of patients with the second type of diabetes mellitus increases. There can be several causes to this disease. One of them is an unhealthy lifestyle and along with it bad eating habits, very little or none of any physical activities and from this deriving obesity. The aim of my thesis was to discover the lifestyle of seniors with the second type of diabetes mellitus, with the consideration of their nutrition and physical activities. 1. The following research questions were...

  18. Koncovka -ě v dativu a lokálu plurálu desubstantivních adjektiv typu otcův a matčin

    Czech Academy of Sciences Publication Activity Database

    Kopáčková, Lucie

    2017-01-01

    Roč. 2017, č. 16 (2017), s. 27-34 ISSN 1804-137X Institutional support: RVO:68378092 Keywords : desubstantive adjective * type otcův * type matčin * suffix -ův * suffix -in * mixed declination * the ending -ě * dative plural * local plural Subject RIV: AI - Linguistics OBOR OECD: Linguistics

  19. Prospects of VVER-SKD reactor in a closed fuel cycle

    OpenAIRE

    Glebov, A.P.; Klushin, A.V.; Yu.D. Baranaev

    2015-01-01

    At the new centure's begin eight countries with developed nuclear power industry took part under the aegis of the IAEA in research of innovative nuclear reactors and fuel cycles to choose a nuclear power system with fast reactors based on a closed fuel cycle (CFC) and to perform joint R&D in this direction. An agreement was reached on the use of based on proven technologies CNFC-FR (Closed Nuclear Fuel Cycles and Fast Reactors), as a reference system for common assessment. Common principle...

  20. Primary coolant technology in VVER/PWR units. Experience with preconditioning, decontamination and recontamination

    Energy Technology Data Exchange (ETDEWEB)

    Vonkova, Katerina [Nuclear Research Institute, Rez (Czech Republic); Kysela, Jan

    2012-01-15

    For the latest Czech and Slovak stations commissioned (Temelin and Mochovce) a modified hot functional test (HFT) chemistry was developed in the NRI Rez. Chromium rich surface layer formed due to modified HTF chemistry ensures lower corrosion rates and radiation field formation. Long term operation experience from both nuclear power plants are discussed in this paper. Radiation field, occupational radiation exposure and corrosion layers evolution during the first 10 years of operation are compared and presented. The operation experience from all above mentioned units showed a low level of corrosion products in the primary system as well as low dose rates. Second part of the paper deals with radiation fields that exist in nuclear power plants primarily due to the deposition of radioisotopes on the surfaces of primary components after decontamination. Large-scale crud deposition on fuel surface resulted in cases NPP Loviisa, Paks and Novovoronezh after steam generators decontamination. After decontamination higher corrosion products release occurs followed by subsequent higher radiation fields. Actual in-pile loop tests carried at the Nuclear Research Institute (NRI) Rez are focused on the study of surface preconditioning - similar to HFT chemistry - after decontamination. Effects of the decontamination on deposition formation on primary circuit surfaces are investigated under steam generator operating conditions with the model device which contains heat exchanger tube. (orig.)

  1. The analysis of normative requirements to materials of VVER components, basing on LBB concepts

    Energy Technology Data Exchange (ETDEWEB)

    Anikovsky, V.V.; Karzov, G.P.; Timofeev, B.T. [CRISM Prometey, St. Petersburg (Russian Federation)

    1997-04-01

    The paper demonstrates an insufficiency of some requirements native Norms (when comparing them with the foreign requirements for the consideration of calculating situations): (1) leak before break (LBB); (2) short cracks; (3) preliminary loading (warm prestressing). In particular, the paper presents (1) Comparison of native and foreign normative requirements (PNAE G-7-002-86, Code ASME, BS 1515, KTA) on permissible stress levels and specifically on the estimation of crack initiation and propagation; (2) comparison of RF and USA Norms of pressure vessel material acceptance and also data of pressure vessel hydrotests; (3) comparison of Norms on the presence of defects (RF and USA) in NPP vessels, developments of defect schematization rules; foundation of a calculated defect (semi-axis correlation a/b) for pressure vessel and piping components: (4) sequence of defect estimation (growth of initial defects and critical crack sizes) proceeding from the concept LBB; (5) analysis of crack initiation and propagation conditions according to the acting Norms (including crack jumps); (6) necessity to correct estimation methods of ultimate states of brittle an ductile fracture and elastic-plastic region as applied to calculating situation: (a) LBB and (b) short cracks; (7) necessity to correct estimation methods of ultimate states with the consideration of static and cyclic loading (warm prestressing effect) of pressure vessel; estimation of the effect stability; (8) proposals on PNAE G-7-002-86 Norm corrections.

  2. TACIS 91: Application of leak-before-break concept in VVER 440-230

    Energy Technology Data Exchange (ETDEWEB)

    Bartholome, G.; Faidy, C.; Franco, C. [and others

    1997-04-01

    The applicability of the leak-before-break (LBB) concept for primary piping in the first generation of WWER type plants in Russia is investigated. The procedures for LBB behavior used in France and Germany are applied, and the evaluation is discussed within the framework of the European Technical Assistance for the Community of Independent States (TACIS) project. Emphasis is placed on experimental validation of national and international engineering practice for evaluating and optimizing existing installations. Design criteria of WWER plants are compared to western standard design.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

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

  4. Use of gadolinium burnable absorbers in VVER Type Reactors. Validation of WIMS-D/4 code; Empleo del gadolinio como absorbente quemable en los reactores nucleares VVER. Validacion del codigo WIMS-D/4

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez Cardona, Caridad M.; Guerra Valdes, Ramiro; Lopez Aldama, Daniel [Centro de Tecnologia Nuclear, La Habana (Cuba)

    1996-07-01

    Burnable absorbers are not used in current operating WWERs, but in order to optimize the fuel cycle and enhance operational safety, one should also introduce gadolinium or a similar burnable absorber in these reactors. For this purpose adequate tools for properly calculating local effects in hexagonal geometries should be developed and validated. The present gives main results in validating the WIMS-D/4 lattice code for Gd burnable absorber bearing WWER lattices. To validate the code experimental and calculational benchmarks proposed in a IAEA Coordinated Research Program were solved. A code system for the optimization of the Gd axial distribution in a WWER reactor was developed and it also presented here. (author)

  5. Validation of finite difference core diffusion calculation methods with FEM and NEM for VVER-1000 MWe reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); RPDD, Central Complex, BARC, Mumbai - 400085 (India); Singh, T. [Reactor Physics and Nuclear Engineering Section, Reactor Group, BARC, Mumbai (India); Pal, U.; Karthikeyan, R. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); Sundaram, G. [Nuclear Safety Group, KK-NPC, Mumbai (India)

    2006-07-01

    India is developing several in-house fuel management codes for the design evaluation of WER-1000 M We reactors, being built at Kudankulam, Tamil Nadu in collaboration with Russian Federation. A lattice burnup code EXCEL provides the few group lattice parameters of various fuel assembly types constituting the core. The core diffusion analyses have been performed by two methods. In the first method the entire fuel assembly is treated as a single homogenized cell. Each fuel assembly cell is divided into 6n{sup 2} triangles, where 'n' is the number of uniform divisions on a side of the hexagon. Regular triangular meshes are used in the active core as well as in surrounding reflector regions. This method is incorporated in the code TRIHEXFA. In the second method a pin by pin description of the core is accomplished by considering the few group lattice parameters generated by EXCEL code for various fuel and non-fuel cells in each fuel assembly. Regular hexagonal cells of one pin pitch are considered in the core and reflector regions. This method is incorporated in HEXPIN code. Both these codes use centre mesh finite difference method (FDM) for regular triangular or hexagonal meshes. It is well known that the large size of the WER fuel assembly, the zigzag structure of the core-baffle zone, the distribution of water tubes of different diameter in this baffle zone and the surrounding steel and water layers of different thickness, all lead to a very complex description of the core-reflector interface. We are analyzing the WER core in fresh state by two other approaches to obtain independent benchmark reference solutions. They are finite element method (FEM) and nodal expansion method (NEM). The few group cross sections of EXCEL are used in the FEM and NEM analyses. The paper would present the comparison of the results of core followup simulations of FD codes with those of FEM and NEM analyses. (authors)

  6. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P. [Nuclear Research Inst. Rez (Switzerland)

    1995-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  7. Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

    CSIR Research Space (South Africa)

    Hirschberg, G

    1999-03-01

    Full Text Available Ag deposition is needed, which may facilitate the elaboration of more e cient surface prevention and/or decontamination procedures. As demonstrated above, corrosion and contamina- tion processes in the primary cooling circuit of PWRs are essentially interrelated...: the contaminant isotopes are mostly corrosion products activated in the reactor core, and the contamination takes place on surfaces which were modi?ed by the corrosion. Also, the two counter measures (decontamination and corrosion-prevention) are connected to each...

  8. Využití databázových systémů v sw balících typu ERP/ERP II

    OpenAIRE

    Vašek, Martin

    2010-01-01

    This thesis is concerning with problems of using database systems in ERP / ERP II software packages. The goal is to define position of ERP / ERP II systems in the Information System market. With this topic are also connected characteristics of database systems and definition of their specific position towards ERP / ERP II solutions. Except classical solutions, when the whole Information System is situated "inside" a company, there are also analyzed new attitudes, which respect external provid...

  9. Mechanismus sekrece adenylát cyklázového toxinu Bordetella pertussis pomocí sekrečního aparátu typu I (TISS)

    OpenAIRE

    Klímová, Nela

    2013-01-01

    Type I secretion system in Gram-negative bacteria translocates proteins from the cytoplasm to the extracellular medium in a single step across both membranes. The membrane-spanning channel is made up of just three proteins - an ATPase in the inner membrane, a membrane fusion protein and a specific outer membrane protein. This work provides a summary of current knowledge concerning the structure of the secretion system, as well as the assembly of the trans-envelope complex and the mechanism of...

  10. Radionuclide inventories in the discharged fuels of PHWR-220, BWR-160, VVER-1000 and the conceptual ATBR-600 reactors - A case study

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Usha [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: ushapal@barc.gov.in; Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: vjagan@barc.gov.in

    2008-10-15

    Radionuclides content in the discharged fuel of the conceptual thorium breeder reactor ATBR-600 has been assessed and compared against other thermal power reactors considered in Indian nuclear power programme. The contribution of actinides and the fission products inventories in the discharged fuels are separately estimated and assessed. The ATBR-600 reactor is suggested for closed fuel cycle option. The relatively large presence of the unspent plutonium would in fact be recycled. Nonetheless, the data has been presented in the event of operating ATBR-600 like other present day power reactors in a once through fuel cycle mode.

  11. Main results of study on the interaction between the corium melt and steel in the VVER-1000 reactor vessel during a severe accident performed under the MASCA project

    Science.gov (United States)

    Asmolov, V. G.; Zagryazkin, V. N.; Tsurikov, D. F.; Vishnevsky, V. Yu.; D'Yakov, Ye. K.; Kotov, A. Yu.; Repnikov, V. M.

    2010-12-01

    The interactions that take place in the corium melt in the reactor vessel in the case of a severe accident at a nuclear power plant were investigated in accordance with the MASCA international program. Results of the interaction between the oxide melt and iron (steel), partition of the main components [U, Zr, Fe (stainless steel)] between the oxide and the metal phases of the melt, partition of low-volatile simulators of fission products between the phases of the stratified core melt pool, and impact of the oxidizing atmosphere on the melt stratification are presented. The results obtained were used for prediction of thermodynamic properties of the melts belonging to the U-Zr-Fe-O system.

  12. Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

    Science.gov (United States)

    Proskuryakov, K. N.; Fedorov, A. I.; Zaporozhets, M. V.

    2015-08-01

    The accident at the Japanese Fukushima Daiichi nuclear power plant (NPP) caused by an earthquake showed the need of taking further efforts aimed at improving the design and engineering solutions for ensuring seismic resistance of NPPs with due regard to mutual influence of the dynamic processes occurring in the NPP building structures and process systems. Resonance interaction between the vibrations of NPP equipment and coolant pressure pulsations leads to an abnormal growth of dynamic stresses in structural materials, accelerated exhaustion of equipment service life, and increased number of sudden equipment failures. The article presents the results from a combined calculation-theoretical and experimental substantiation of mutual amplification of two kinds of external periodic loads caused by rotation of the reactor coolant pump (RCP) rotor and an earthquake. The data of vibration measurements at an NPP are presented, which confirm the predicted multiple amplification of vibrations in the steam generator and RCP at a certain combination of coolant thermal-hydraulic parameters. It is shown that the vibration frequencies of the main equipment may fall in the frequency band corresponding to the maximal values in the envelope response spectra constructed on the basis of floor accelerograms. The article presents the results from prediction of conditions under which vibroacoustic resonances with external periodic loads take place, which confirm the occurrence of additional earthquake-induced multiple growth of pressure pulsation intensity in the steam generator at the 8.3 Hz frequency and additional multiple growth of vibrations of the RCP and the steam generator cold header at the 16.6 Hz frequency. It is shown that at the elastic wave frequency equal to 8.3 Hz in the coolant, resonance occurs with the frequency of forced vibrations caused by the rotation of the RCP rotor. A conclusion is drawn about the possibility of exceeding the design level of equipment vibrations under the effect of external periodic loads caused by an earthquake when the vibration frequency of the reactor plant main equipment and the frequency of elastic waves fall in the frequency band corresponding to the maximal values of envelope response spectra.

  13. 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.

  14. RISKAUDIT Report no. 7, Vol. 2: Safety evaluation of VVER 440/213 and VVER 1000/320 reactors in Rovno NPP Units 1, 2 and 3. Final Report by AIB-Vincotte Nuclear, CIEMAT, ANPA, GRS, IPSN, AEA-T

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-15

    The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The second part of the document covers the following aspects of the report: accident analysis; systems analysis; plant operation; operating experience feedback; radio protection and health; probabilistic safety assessment; summary and future plans.

  15. RISKAUDIT Report no. 7, Vol. 1: Safety evaluation of VVER 440/213 and VVER 1000/320 reactors in Rovno NPP Units 1, 2 and 3. Final Report by AIB-Vincotte Nuclear, CIEMAT, ANPA, GRS, IPSN, AEA-T

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-15

    The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The first part of the document covers the following aspects of the report: core design and fuel management; pressurized components; electrical supply; instrumentation and control; containment; internal events; site conditions and external events.

  16. Skew Bending of Aircraft Fuselage Panels with “L” and “C” Stringers Mounted by Hybrid Joint / Ukośne Zginanie Poszycia Samolotu Z U Sztywnieniami Typu “L” I “C”, Mocowanymi Za Pomocą Złącza Hybrydowego

    Directory of Open Access Journals (Sweden)

    Sadowski T.

    2015-12-01

    Full Text Available A section of fuselage skin with dimension 30 x 200 mm was subjected to numerical study and loaded by skew bending (Fig. 3. The thickness of the skin was 0,6 mm, the length of a leg of an angle “L” profile stringer was 12 mm with 1mm thickness. The angle of inclination α of the load plane to the skin plane varies in the range from 10° to 90° with 10° increment. The elastic - plastic material model of D16T aluminum alloy was used in simulations of the fuselage skin as well as for “L” and “C” profile stringers. In the material model description damage of aluminum alloy was taken into account. An adhesive layer with thickness of 0,1mm was modeled using cohesive elements with the failure mode depending on the shear strength and the tensile strength.

  17. Souvislost mezi řízením vztahů k zákazníkům (CRM a tržní orientací (MO a vliv velikosti podniku a typu trhu na CRM a MO

    Directory of Open Access Journals (Sweden)

    Daniela Frejková

    2014-03-01

    Full Text Available Purpose of the article: This paper is concerned with Market Orientation (MO and Customer Relationship Management (CRM. These two topics have been frequently explored in the field of corporate management and marketing. Nevertheless, these two concepts are still analyzed separately in the literature. This article explains why these concepts are interdependent and sets the degree of dependence of these concepts. It also finds out whether the level of MO and CRM is dependent on company size or the type of market. Methodology/methods: This article has been prepared based on the analysis of secondary and primary sources. The primary research was conducted on a sample of 29 completed questionnaires provided by firms from the aerospace field in the Czech Republic. The level of CRM and MO was determined for each company and a statistical verification was conducted. Scientific aim: One aim of this article is to reveal the interconnections between MO and CRM. Other aim is to determine whether CRM and MO are affected by the size of the company and whether they depend on the type of market (business-to-business, i.e. B2B or business-to-customer, i.e. B2C. Findings: Findings of this article are new information in this area. The data strongly support the proposition that MO is interconnected with CRM, while no dependence on the size of the company or the type of market has been confirmed. Conclusions: This research supports the opinion that MO and CRM are appropriate for each of the researched type of the company (under certain conditions. The findings must be considered within the limitations of this study. Conclusions for the whole business may be drawn after the comparison of experiences across business sectors from different countries.

  18. The Influence of Geometrical Parameters in Socket - Pin Connections on the Value of Opening Force / Wpływ Parametrów Geometrycznych W Połączeniach Typu Gniazdo - Trzpień Na Wartość Siły Otwierającej

    Directory of Open Access Journals (Sweden)

    Sadowski T.

    2015-12-01

    Full Text Available The paper presents an analysis of the influence of a number of technological aspects of both the socket and the pin on the value of the force required for joint disconnection. A number of numerical simulations were made in Abaqus program to examine effects of such parameters as: presence of an interference fit, use of spherical latches, application of different rigidity of the pin by making cuts with variable width and length, use of different angles of inclination of the working part of the connection. Models of different simple joints presented in this work, can also operate in large structures forming panels of aircraft structures. For this purpose one of the analyzed geometry of the connection was applied to create a 3-D panel model of the structural element in CAD - SolidWorks program. All analysed models with different geometries were subjected to simulation of opening process. The corresponding critical forces were estimated for the beginning of the failure process. The detailed discussion of all model parameters was included to specify their influence on the whole disconnection of joints. It should be noted that aerospace structures work under complex loading states and further numerical studies are required to extend the presented results.

  19. Subjektivní pohled na kvalitu života diabetiků prvního typu léčených kontinuální subkutánní inzulínovou infuzí.

    OpenAIRE

    MACHYÁNOVÁ, Klára

    2007-01-01

    The work is interesting in quality of life of patients with type 1 diabetes treated by CSII. The target of work is recognize subjective view on a quality of life of diabetics and appraising domain of his life which is the weakest from them.

  20. Wpływ estrogenów na stężenie N-terminalnego propeptydu kolagenu typu I w płynie uzyskanym z przestrzennych hodowli ludzkich fibroblastów powięzi łonowo-cewkowej prowadzonych na siatkach polipropylenowych stosowanych w uroginekologii operacyjnej

    Directory of Open Access Journals (Sweden)

    Jacek Tomaszewski

    2010-02-01

    Full Text Available Objectives: Polypropylene meshes are widely used for surgical treatment of stress urinary incontinence(SUI and pelvic organ prolapse (POP. Synthesis and deposition of collagen induced by an inserted implant arelargely controlled by oestrogens. The aim of the study was to assess the rate of collagen type I (Col I synthesisby pubo-cervical fascia (PCF fibroblasts cultured with mono- or multifilament polypropylene meshes in thepresence of oestrogens. Material and Methods: Specimens of PCF were obtained during a surgical procedure from a 56-year-old womansuffering from SUI and POP. Fibroblasts were cultured with mono- or multifilament meshes and exposed to17β-oestradiol, oestriol or phytoestrogen daidzein. The cultures were run for 216 hrs and the media were replaced every 72 hrs. Procollagen type 1 N-terminal propeptide (PINP, a marker of Col I biosynthesis, was assessedin culture media by radioimmunoassay. Results: The biosynthesis of Col I was more abundant in the presence of monofilament than multifilamentmeshes. Fibroblasts exposed to oestriol or daidzein produced more Col I than those treated with oestradiol,regardless of the mesh applied. In the presence of monofilament mesh the rate of Col I synthesis induced byoestriol and daidzein increased persistently until the end of the experiment, whereas the peak concentration ofPINP in cultures treated with oestradiol was observed between 72 hrs and 144 hrs. In the presence of multifilamentmesh the rate of Col I production dropped after 144 hrs in all cultures. Conclusions: PCF fibroblasts produce more Col I when cultured on monofilament than on multifilamentmesh. This process may be enhanced by oestriol and phytoestrogens.

  1. Effectiveness of the neutron-shield nanocomposites for a dual-purpose cask of Bushehr's Water–Water Energetic Reactor (VVER 1000 nuclear-power-plant spent fuels

    Directory of Open Access Journals (Sweden)

    Mahdi Rezaeian

    2017-10-01

    Full Text Available In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP code. The dose rate for the dual-purpose cask utilizing the recently developed materials of epoxy/clay/B4C and epoxy/clay/B4C/carbon fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of epoxy/clay/B4C instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.

  2. The criterion for blanking-off heat-transfer tubes in the steam generators at VVER-based nuclear power plants based on the results of eddy-current examination

    Science.gov (United States)

    Lunin, V. P.; Zhdanov, A. G.; Chegodaev, V. V.; Stolyarov, A. A.

    2015-05-01

    The problem of defining the criterion for blanking off heat-transfer tubes in the steam generators at nuclear power plants on the basis of signals obtained from the standard multifrequency eddy-current examination is considered. The decision about blanking off one or another tube is presently made with reference to one parameter of the relevant signal at the working frequency, namely, with reference to its phase, which directly depends on the depth of the flaw being detected, i.e., a crack in the tube. The crack depth equal to 60% of the tube wall thickness is regarded to be the critical one, at which a decision about withdrawing such a tube out from operation (blanking off) must be taken. However, since mechanical tensile rupture tests of heat-transfer tubes show the possibility of their further use with such flaws, the secondary parameter of the signal, namely, its amplitude, must be used for determining the blanking-off criterion. The signals produced by the standard flow-type transducers in response to flaws in the form of a longitudinal crack having the depth and length within the limits permitted by the relevant regulations were calculated using 3D finite-element modeling. Based on the obtained results, the values of the eddy-current signal amplitude were determined, which, together with the signal phase value, form a new amplitude-phase criterion for blanking off heat-transfer tubes. For confirming the effectiveness of this technique, the algorithm for revealing the signal indications satisfying the proposed amplitude-phase criterion was tested on real signals obtained from operational eddy-current examination of the state of steam generator heat-transfer tubes carried out within the framework of planned preventive repair.

  3. The Valve Gear Systems Timing Parameters Identification for Marine Diesel Engines Diagnostics / Identyfikacja Parametrów Układów Rozrządu Silników Okrętowych Dla Potrzeb Diagnostyki

    OpenAIRE

    Lus Tomasz

    2015-01-01

    W referacie przedstawiono wyniki badań, których celem było opracowanie metody diagnozowania okrętowych szybkoobrotowych silników tłokowych. Polska Marynarka Wojenna eksploatuje znaczną liczbę silników tego typu również na okrętach podwodnych. Silniki tego typu nie posiadają zaworów indykatorowych, co komplikuje możliwość oceny procesu spalania i ich stanu technicznego. Akademia Marynarki Wojennej w Gdyni od lat prowadzi badania związane z rozwojem metod diagnozowania okrętowych tłokowych siln...

  4. Development of codes and KASKAD complex

    Energy Technology Data Exchange (ETDEWEB)

    Lizorkin, Mikhail; Gordienko, P.V.; Kalugin, M.A.; Kotsarev, A.V.; Oleksyuk, D.A. [Nuclear Research Center ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A brief description of the development history of engineering computer codes for VVER reactor calculations, which have been developed in the VVER physics department of NRC Kurchatov institute, since the end of 1950 s till now, are given in the report. The modern status of the basic codes of the Kurchatov institute for VVER fuel loadings design and safety analysis (TVS-M, KASKAD-BIPRPERMAK, BIPR8, PERMAK-3D, SC1, TOPRA, ATHLET/ BIPR-VVER, TIGR-1, etc.) is characterized from various points of view. Some specific problems which are shown during designing of fuel cycles, optimisation of fuel and technologies will be commented. They demand decisions within the complex of programs and, accordingly, demand the further development of a complex or working out and inclusion in a complex of new programs.

  5. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Directory of Open Access Journals (Sweden)

    Luben Sabotinov

    2010-01-01

    Full Text Available This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER. The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model, represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

  6. Modelling studies of horizontal steam generator PGV-1000 with Cathare

    Energy Technology Data Exchange (ETDEWEB)

    Karppinen, I. [VTT Energy, Espoo (Finland)

    1995-12-31

    To perform thermal-hydraulic studies applied to nuclear power plants equipped with VVER, a program of qualification and assessment of the CATHARE computer code is in progress at the Institute of Protection and Nuclear Safety (IPSN). In this paper studies of modelling horizontal steam generator of VVER-1000 with the CATHARE computer code are presented. Steady state results are compared with measured data from the fifth unit of Novovoronezh nuclear power plant. (orig.). 10 refs.

  7. Overall Plan for Physics Outlining Steps Necessary for Insertion of the LTA and Operation Using a 1/3 MOX Loaded Core

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-04-09

    Document issued according to Work Release KI-WR04RTP. P. 00-1 describes physics tasks that are included in the current version of ''Roadmap.Level 2'' concerning Reactor tasks of Weapon-grade plutonium disposition problem for VVER-1000. On this base the objective is to identify the physical tasks in FY2000 and in future as a part of global activities on weapon-grade MOX fuel introduction into VVER-1000.

  8. Developing the concept of maintenance and repairs in projects of power units for new-generation nuclear power stations

    Science.gov (United States)

    Gurinovich, V. D.; Yanchenko, Yu. A.

    2012-05-01

    Results from conceptual elaboration of individual requirements for the system of maintenance and repairs that must be implemented in the projects of new-generation nuclear power stations are presented taking as an example the power unit project for a nuclear power station equipped with a standard optimized VVER reactor with enhanced information support (the so-called VVER TOI reactor). Implementation of these concepts will help to achieve competitiveness of such nuclear power stations in the domestic and international markets.

  9. Stress corrosion (Astm G30-90 standard) in 08x18H10T stainless steel of nuclear fuel storage pool in WWER reactors; Corrosion bajo esfuerzo (Norma ASTM G30-90) en acero inoxidable 08x18H10T de piscinas de almacenamiento de combustible nuclear en reactores V.V.E.R

    Energy Technology Data Exchange (ETDEWEB)

    Herrera, V.; Zamora R, L. [Centro de Estudios Aplicados al Desarrollo Nuclear (Cuba)

    1997-07-01

    At the water storage of the irradiated nuclear fuel has been an important factor in its management. The actual pools have its walls covered with inoxidable steel and heat exchangers to dissipate the residual heat from fuel. It is essential to control the water purity to eliminate those conditions which aid to the corrosion process in fuel and at related components. The steel used in this research was obtained from an austenitic inoxidizable steel standardized with titanium 08x18H10T (Type 321) similar to one of the two steel coatings used to cover walls and the pools floor. the test consisted in the specimen deformation through an U ply according to the Astm G30-90 standard. The exposition of the deformed specimen it was realized in simulated conditions to the chemical regime used in pools. (Author)

  10. Recent results on the RIA test in IGR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  11. Fatigue flaw growth assessment and inclusion of stratification to the LBB assessment

    Energy Technology Data Exchange (ETDEWEB)

    Samohyl, P.

    1997-04-01

    The application of the LBB requires also fatigue flaw growth assessment. This analysis was performed for PWR nuclear power plants types VVER 440/230, VVER 440/213c, VVER 1000/320. Respecting that these NPP`s were designed according to Russian codes that differ from US codes it was needed to compare these approaches. Comparison with our experimental data was accomplished, too. Margins of applicability of the US methods and their modifications for the materials used for construction of Czech and Slovak NPP`s are shown. Computer code accomplishing the analysis according to described method is presented. Some measurement and calculations show that thermal stratifications in horizontal pipelines can lead to additive loads that are not negligible and can be dangerous. An attempt to include these loads induced by steady-state stratification was made.

  12. Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system

    Energy Technology Data Exchange (ETDEWEB)

    Gurin, Andrey V. [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation); Alekseev, P.N.

    2017-09-15

    This paper presents a study of scenarios of transition to a closed fuel cycle in the system of nuclear power, built basing on resource availability requirements at the stage of full life-cycle reactors. Conventionally, there are three main scenarios for the development of nuclear energy: with VVER reactors operating in an open fuel cycle; with VVER reactors operating in a closed fuel cycle; and co-operating VVER and BN, operating in a closed fuel cycle. For the considered scenarios, a quantitative estimation of change in time of material balances were performed, including spent fuel balance, balance of plutonium, reprocessed and depleted uranium, radioactive waste, and the analysis of the fuel component of the cost of electricity.

  13. Książki zabawki jako typ książki dla dzieci

    OpenAIRE

    Sołtysiewicz, Anna

    2015-01-01

    Artykuł został poświęcony gatunkowi książek zabawek. Przedstawiłam w nim definicje i funkcje tego typu publikacji. Następnie analizie statystycznej poddałam informacje zawarte w bazie danych książek zabawek. Wyniki tej analizy posłużyły do opisu sytuacji na polskim rynku książek zabawek. Ustaliłam wielkość produkcji tego typu pozycji ze względu na rok wydania i wydawców, a także najpopularniejsze serie wydawnicze. Określiłam, którzy autorzy i ilustratorzy są twórcami największej liczby książe...

  14. Styrene Oxidation by Copper(II Complexes Salen-Type Encapsulated into Nay Zeolite

    Directory of Open Access Journals (Sweden)

    Kuźniarska-Biernacka I.

    2013-12-01

    Full Text Available Osadzenie kompleksu miedzi(II z zasadą Schitta typu salen na zeolicie typu NaY zostało prze powadzone za pomocą dwóch metod „flexible ligand" i „in situ”. Katalityczne właściwości otrzymanego kompleksu oraz jego heterogenizowanych analogów badano w reakcji utleniania styrenu w obecności TBHP. Jako rozpuszczalnik stosowano acetonitryl. Testowane katalizatory wykazują średnią aktywność katalityczną z tendencją wytwarzania aldehy du benzoesowego. Oba heterogenizowane katalizatory mogą być wykorzystywane ponownie bez utraty aktywności katalitycznych.

  15. Structural seismic upgrading of NPPs in Czech and Slovak republics

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

    Several Nuclear Power Plants of the VVER type has been constructed during the past years in former Czechoslovak Republic. Some of them has been already put in operation and some of them are under construction. Nuclear Power Plants V1(2 units of VVER 440/230), V2(2 units of VVER 440/213) in Slovak and NPP Dukovany (4 units of VVER 440/213) in Czech republic are in operation. NPP Mochovce (4 units of VVER 440/213) in Slovak and NPP Temelin (4 units reduced now to 2 units VVER 1000) have been already almost completed, but still under construction. All above cited NPPs have not been either explicitly designed against earthquake or the design against earthquake or its input data must be upgraded to be compatible with present requirements. The upgrading of seismic input as well the seismic upgrading of all structures and technological equipments for so many NPPs has involved a lot of comprehensive work in Czech as well as in Slovak republics. The upgrading cannot be completed in a short time and as a rule the seismic upgrading has been usually performed in several steps, beginning with the most important arrangements against seismic hazard. The basic principles and requirements for seismic upgrading has been defined in accordance with the international and particularly with the IAEA recommendations. About the requirements for seismic upgrading of NPPs in Czech and Slovak republics will be reported in other contribution. This contribution is dealing with the problems of seismic upgrading of NNPs civil engineering structures. The aim of this contribution is to point out some specific problems connected firstly with very complicated concept of Versa structures and secondly with the difficult task to increase the structural capacity to the required seismic level. (J.P.N.)

  16. SDR přijímač pro pásmo do 40 MHz

    OpenAIRE

    Vorba, Filip

    2015-01-01

    Tato práce se zabývá popisem konceptů SDR (Software Defined Radio) a zejména návrhem a realizací tohoto typu přijímače pro pásmo do 40 MHz. This project is about designing a new SDR (Software Defined Radio) receiver for band up to 40 MHz. D

  17. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center.

    Energy Technology Data Exchange (ETDEWEB)

    Podlazov, L. N.

    1998-07-29

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions.

  18. Fourth international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Tuomisto, H. [ed.] [IVO Group, Vantaa (Finland); Purhonen, H. [ed.] [VTT, Espoo (Finland); Kouhia, V. [ed.] [Lappeenranta Univ. of Technology (Finland)

    1997-12-31

    The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

  19. The Valve Gear Systems Timing Parameters Identification for Marine Diesel Engines Diagnostics / Identyfikacja Parametrów Układów Rozrządu Silników Okrętowych Dla Potrzeb Diagnostyki

    Directory of Open Access Journals (Sweden)

    Lus Tomasz

    2015-09-01

    Full Text Available W referacie przedstawiono wyniki badań, których celem było opracowanie metody diagnozowania okrętowych szybkoobrotowych silników tłokowych. Polska Marynarka Wojenna eksploatuje znaczną liczbę silników tego typu również na okrętach podwodnych. Silniki tego typu nie posiadają zaworów indykatorowych, co komplikuje możliwość oceny procesu spalania i ich stanu technicznego. Akademia Marynarki Wojennej w Gdyni od lat prowadzi badania związane z rozwojem metod diagnozowania okrętowych tłokowych silników spalinowych. W ostatnich latach prowadzone są prace nad metodami diagnozowania szybkoobrotowych silników okrętowych bazującymi na analizie obwiedni przyspieszeń drgań generowanych przez układy rozrządu zaworowego i układ paliwowy. W referacie zaprezentowano wybrane wyniki badań prowadzonych na silnikach Mercedes-Maybach typu MB820 stosowanych na okrętach podwodnych klasy Kobben.

  20. ALM – nowatorskie metryki wskaźników wpływu w publikacjach naukowych

    Directory of Open Access Journals (Sweden)

    Martin Fenner

    2015-12-01

    Full Text Available Cytowania przytaczane w literaturze światowej, a ostatnio statystyki wykorzystania materiału naukowego dostarczające informacji na temat zainteresowania i oddźwięku, jakie zdobywają publikacje naukowe, nadal pozostają ważnymi wskaźnikami wpływu w ocenie publikacji. Ukazują one bowiem widoczny wkład danej pracy naukowej w rozwój nauki i badań. Public Library of Science (PLOS, platforma projektu typu non profit oferująca dostęp do czasopism naukowych na zasadzie licencji wolnej dokumentacji oraz wydawca publikacji typu open access, była jednym z pierwszych tego typu projektów, w którym wprowadzono możliwość analizy innowacyjnych metryk, tzw. article-level metrics (ALM, tj. badania wpływu publikacji naukowych na poziomie artykułu, a nie czasopisma, z uwzględnieniem zmian, jakie zaszły w komunikacji naukowej wraz z rozwojem sieci. Badania uzupełniane są danymi kontekstowymi świadczącymi o zakresie społecznej uwagi poświęcanej danej pracy naukowej, uzyskiwanymi w czasie rzeczywistym z naukowych portali społecznościowych lub z alternatywnych metryk, takich jak blogi naukowe, zakładki, media społecznościowe.

  1. Fulltext PDF

    Indian Academy of Sciences (India)

    An International Benchmark Exercise,. NEA/OECD (2000). [7] Benchmark on the Venus-2 Mox Core Measurements, NEA/NSC/DOC (2000) 7. [8] A D Klimov et al, System analysis of nuclear safety of VVER reactor with Mox fuel,. Proc. Int. Conf. Mathematics and Computation, Supercomputing, Reactor Physics and. Nuclear ...

  2. Mission Fuel Kinetics Input and RELAP-like Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-09-28

    In this document issued according to ''Work Release 02. P. 99-4b'' the neutronics parameters intended for use in 1-point kinetics RELAP model are presented. They are obtained for equilibrium 30% MOX fueled core of VVER-1000 containing boron burnable poison rods.

  3. Initial data on protective containment for safety analysis of NPP with WWER-1000 Preparation

    Directory of Open Access Journals (Sweden)

    Gubeladze Oleg

    2017-01-01

    Full Text Available In the presented article the nuclear safety issues with VVER-1000 reactor are considered. The study element is the protective containment (PC, the most important function of which is localization and retention of radioactive substances within the accident localization zone. The example of possible unregulated destructive forcing (UDF on the PC for the construction and installation work period is given.

  4. Third international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

  5. MEASUREMENTS OF THE CONFINEMENT LEAKTIGHTNESS AT THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-08-01

    This is the final report on the INSP project entitled, ``Kola Confinement Leaktightness'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.1. This project was initiated in February 1993 to assist the Russians to reduce risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Units 1 and 2, through upgrades in the confinement performance to reduce the uncontrolled leakage rate. The major technical objective of this-project was to improve the leaktightness of the Kola NPP VVER confinement boundaries, through the application of a variety of sealants to penetrations, doors and hatches, seams and surfaces, to the extent that current technology permitted. A related objective was the transfer, through training of Russian staff, of the materials application procedures to the staff of the Kola NPP. This project was part of an overall approach to minimizing uncontrolled releases from the Kola NPP VVER440/230s in the event of a serious accident, and to thereby significantly mitigate the consequences of such an accident. The US provided materials, application technology, and applications equipment for application of sealant materials, surface coatings, potting materials and gaskets, to improve the confinement leaktightness of the Kola VVER-440/23Os. The US provided for training of Russian personnel in the applications technology.

  6. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  7. EBO feed water distribution system, experience gained from operation

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O. [Energovyzkum, Brno (Switzerland); Schmidt, S.; Mihalik, M. [Atomove Elektrarne Bohunice, Jaslovske Bohunice (Switzerland)

    1997-12-31

    Advanced feed water distribution systems of the EBO design have been installed into steam generators at Units 3 and 4 of the NPP Jaslovske Bohunice (VVER 440). Experiences gained from the operation of steam generators with the advanced feed water distribution systems are discussed in the paper. (orig.). 4 refs.

  8. On the equilibrium isotopic composition of the thorium–uranium–plutonium fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Marshalkin, V. Ye., E-mail: marshalkin@vniief.ru; Povyshev, V. M. [Russian Federal Nuclear Center—All-Russian Scientific Research Institute of Experimental Physics (Russian Federation)

    2016-12-15

    The equilibrium isotopic compositions and the times to equilibrium in the process of thorium–uranium–plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

  9. Twardzina ograniczona pęcherzowa – opis przypadku

    OpenAIRE

    Elżbieta Meszyńska; Ligia Brzezińska-Wcisło; Beata Bergler-Czop

    2011-01-01

    Wprowadzenie: Pęcherze pojawiające się w obrębie zmian typu twardzinyograniczonej są rzadko opisywane. Etiologia ich powstawaniajest wieloczynnikowa. Najczęściej rozważany jest ucisk blaszek twardzinowychna naczynia limfatyczne. Twardzina pęcherzowa wymagaróżnicowania z liszajem twardzinowym i zanikowym, w którym u częścichorych obserwuje się występowanie pęcherzy. Cel pracy: Przedstawienie przypadku 41-letniej kobiety z twardzinąograniczoną (morphea), u której w 15. roku trwania choroby wyst...

  10. Přizpůsobení platformy LLVM pro mikroprocesor Motorola 68000

    OpenAIRE

    Blahož, Vladimír

    2010-01-01

    Bakalářská práce popisuje obecnou problematiku překladačů, seznamuje čtenáře s platformou Low-Level Virtual Machine a možnostmi její modifikace. Dále se zabývá principy architektury typu Motorola 68000 a implementací podpory její instrukční sady pro platformu LLVM. This bachelor's thesis deals with general questions of compilers, describes the Low-Level Virtual Machine platform and its modification options. Furthermore it concerns about principles of Motorola 68000 architecture and impleme...

  11. Wpływ temperatury na kinetykę utwardzania wybranych mas ze spoiwami

    Directory of Open Access Journals (Sweden)

    Ł. Jamrozowicz

    2012-12-01

    Full Text Available W artykule zaprezentowano wyniki badań wpływu temperatury osnowy na przebieg procesu wiązania mas samoutwardzalnych.Badaniom poddano 3 masy na bazie: żywicy alkidowej, żywicy fenolowej typu rezolowego (proces -set i żywicy furfurylowej (procesfuranowy. Ponadto wyznaczono kinetykę procesu utwardzania i określono żywotność masy przy różnych temperaturach osnowy (10oC;20oC; 30oC. Badania prowadzono przy wykorzystaniu techniki ultradźwiękowej.

  12. Wychowanie w szkołach jezuickich okresu staropolskiego

    OpenAIRE

    Kochanowicz, Jerzy

    2010-01-01

    Brak ostatniej strony. Prezentowany artykuł jest próbą zarysowania kompleksowego zagadnienia wychowania w szkole jezuickiej w okresie staropolskim. Z konieczności oferuje jednak spojrzenie uproszczone ze względu na wielość form, w których realizowała się jezuicka oświata (szkoły różnego typu i stopnia, teatr szkolny, sodalicje uczniowskie, instytucje wspomagające: bursy, konwikty, biblioteki etc.), ze względu na jego wzloty i upadki, wreszcie – jego zróżnicowanie geograficzne: szkoły w Wie...

  13. Multiphysics simulation of fast transients with the FINIX fuel behaviour module

    Directory of Open Access Journals (Sweden)

    Ikonen Timo

    2016-01-01

    Full Text Available FINIX is a recently developed fuel behaviour module that is designed to provide “simple but sufficient” descriptions of the most essential fuel behaviour phenomena in multiphysics simulations. In such simulations, it is possible to obtain significant improvement in the feedback to neutronics or thermal hydraulics modelling even with a relatively simple fuel performance model. In this work, FINIX is used as an internal fuel behaviour module both in reactor physics and in reactor dynamics codes to simulate coupled behaviour in fast transient scenarios. With the Monte Carlo reactor physics code Serpent we model a prompt transient in a VVER-1000 pin cell, and with the reactor dynamics code HEXTRAN, a control rod ejection accident in a VVER-440 reactor.

  14. Support calculations for management of PRISE leakage accidents

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranka, L. [Nuclear Power Plants Research Inst. Vuje, Trnava (Slovakia)

    1997-12-31

    Accidents involving primary-to-secondary leakage (PRISE) caused by rupture of one or a few tubes are well known design basis events in both, western and VVER NPPs. Operating experience and in-service inspections of VVER-440 units have demonstrated also the potential for large PRISE leaks in the case of the steam generator (SG) primary collector cover lift-up (Rovno NPP). Without performing any countermeasure for limitation of SG collector cover lift-up, a full opening results in PRISE leak with an equivalent diameter 107 mm. Although this accident was not considered in the original design, this event is usually analysed as DBA too. Different means are available for detection and mitigation of PRISE leakage in NPPs currently in operation (J.Bohunice V-1 and V-2) or under construction (Mochovce) in Slovakia. 8 refs.

  15. Condensation driven water hammer studies for feedwater distribution pipe

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland); Pullinen, J. [IVO Power Engineering Ltd., Vantaa (Finland); Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K. [EDO Gidropress (Russian Federation)

    1997-12-31

    Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.).

  16. Qualification according to PDI's techniques UT EPRI methodology Phased Array for the inspection of vessels of PWR reactor with advanced robotic equipment; Cualificacion segun metodologia PDI de EPRI de te cnicas UT Phased Array para la inspeccion de vasijas de reactor PWR con eq uipos roboticos avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Gonzalez, P.; Fernandez, F.

    2014-07-01

    The techniques and procedures qualified in the program EPRI PDI are directly applicable in plants whose reference code is ASME XI - specifically the Appendix VIII-, mainly USA and countries in which it is established American PWR technology. While countries with reactors in operation technology ABB (Sweden) or type VVER (Finland and Eastern countries) requires a qualification of specific technical type ENIQ, PDI qualification is a valuable reference since it allows to deal with such qualifications with guarantees. (Author)

  17. JPRS Report: Environmental Issues.

    Science.gov (United States)

    1991-12-27

    reactors, Deputy Prime Minister Aleksandur Tomov says. The two oldest units of Soviet VVER-440 design of the mid-1970’s provide between 10 percent and...12 percent of Bulgarian power, Tomov , Minister for Energy and Vice President of the Socialist Party, formerly the Communist Party, told AFP in an...station catastrophe in the Soviet Ukraine, which killed at least 31 people and sent a plume of radiation over much of Europe. Tomov told AFP that he had

  18. Study of steam condensation in SG tubes with large amount of nitrogen to be accumulated

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S.A.; Sitnik, Y.K. [EDO Gidropress, Podolsk (Russian Federation)

    1997-12-31

    The effect of nitrogen during SG heat transfer under SBLOCA conditions have been studied. Depressurization of the primary side leads to release of nitrogen dissolved in the hydroaccumulator water. Nitrogen can accumulate in SGs and affect adversely heat transfer under reflux condenser conditions. The main objective of the study has been to show that nitrogen does not prevent heat transfer in SGs of the VVER-640 which is reactor plant of new generation. (orig.).

  19. Potential consequences in Norway after a hypothetical accident at Leningrad nuclear power plant. Potential release, fallout and predicted impacts on the environment

    Energy Technology Data Exchange (ETDEWEB)

    Nalbandyan, A.; Ytre-Eide, M.A.; Thoerring, H.; Liland, A.; Bartnicki, J.; Balonov, M.

    2012-06-15

    The report describes different hypothetical accident scenarios at the Leningrad nuclear power plant for both RBMK and VVER-1200 reactors. The estimated release is combined with different meteorological scenarios to predict possible fallout of radioactive substances in Norway. For a hypothetical catastrophic accident at an RBMK reactor combined with a meteorological worst case scenario, the consequences in Norway could be considerable. Foodstuffs in many regions would be contaminated above the food intervention levels for radioactive cesium in Norway. (Author)

  20. Seismic assessment and upgrading of nuclear power plants in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Katona, T.; Kostov, M.

    1997-03-01

    The basic findings of the seismic re-qualification programmes going on recently at all VVER plants in Eastern Europe can be summarised. The problems of the seismic safety have to be solved taking into account the general concept of the nuclear safety enhancement of the units. There are cases where the system improvements lead to better and more effective solution of the problem than the structural upgrading. The equipment and piping of the primary system have sufficient capacity. The viscous dampers are considered usually for the upgrading. The equipment anchorage especially the electrical and I and C equipment anchorage have to be upgraded. There are general consideration for replacement of the hydraulic snubbers by viscous dampers in the primary circuit of the VVER 440/V230. The considerations are not only because of the better seismic behaviour but mainly because of the better operational performance. There is relatively good seismic instrumentation at the plants considered. The definition of the scram level of the units not designed for an OBE is an essential problem. More effort needed for the definition of this level on the basis of re-evaluation experience of the plant equipment and after the proper definition of post-earthquake activities. The seismic re-evaluation and re-qualification of the VVER units is a general safety issue in Easter European countries. This rather complex problem can be solved adopting the experience, methods and requirements of western countries and taking into account the design features of the VVER units as well as the as built and as it is conditions. (J.P.N.)

  1. CFD simulation of a four-loop PWR at asymmetric operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jian-Ping; Yan, Li-Ming; Li, Feng-Chen, E-mail: lifch@hit.edu.cn

    2016-04-15

    Highlights: • A CFD numerical simulation procedure was established for simulating RPV of VVER-1000. • The established CFD approach was validated by comparing with available data. • Thermal hydraulic characteristics under asymmetric operation condition were investigated. • Apparent influences of the shutdown loop on its neighboring loops were obtained. - Abstract: The pressurized water reactor (PWR) with multiple loops may have abnormal working conditions with coolant pumps out of running in some loops. In this paper, a computational fluid dynamics (CFD) numerical study of the four-loop VVER-1000 PWR pressure vessel model was presented. Numerical simulations of the thermohydrodynamic characteristics in the pressure vessel were carried out at different inlet conditions with four and three loops running, respectively. At normal stead-state condition (four-loop running), different parameters were obtained for the full fluid domain, including pressure losses across different parts, pressure, velocity and temperature distributions in the reactor pressure vessel (RPV) and mass flow distribution of the coolant at the inlet of reactor core. The obtained results for pressure losses matched with the experimental reference values of the VVER-1000 PWR at Tianwan nuclear power plant (NPP). For most fuel assemblies (FAs), the inlet flow rates presented a symmetrical distribution about the center under full-loop operation conditions, which accorded with the practical distribution. These results indicate that it is now possible to study the dynamic transition process between different asymmetric operation conditions in a multi-loop PWR using the established CFD method.

  2. Calculation of the Novovoronezh Recriticality Experiment with the KARATE-440 code system

    Energy Technology Data Exchange (ETDEWEB)

    Hegyi, György, E-mail: ghegyi@aeki.kfki.hu [MTA KFKI Atomic Energy Research Institute, Budapest (Hungary)

    2011-07-01

    In this paper the results of KARATE-440 calculations on Novovoronezh NPP Recriticality Experiment are presented, the corresponding parameters are analyzed. The simulation of the processes and the comparison of the results with the measurements are of particular interest as these efforts make our code to be validated in a higher level. The KARATE-440 code system has been developed and applied for VVER-440 core analysis during near twenty years, as a close collaboration among the developers and the specialists at the 4 Hungarian nuclear power units. KARATE is now a mature, demonstrated, complete and integrated system of computer codes and procedures that provide full and independent VVER core analysis capabilities. Even if only some well defined states of the experiment were simulated, satisfactory agreement was found between measured and calculated data. The results present evidence that the KARATE- 440 code package can adequately model the reactor states in a wide range of performance parameters and the special core type referred in the experiment so it is acceptable for neutronic analysis of all the VVER-440 NPP's. (author)

  3. Management of primary-to-secondary leaks at Loviisa nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohnsen, B.; Jaenkaelae, K. [IVO International Ltd, Vantaa (Finland)

    1995-12-31

    The Loviisa Nuclear power plant consisting of two VVER-440 type press water reactor units has been in commercial operation since the late 1970`s. Specific features for VVER-440 reactors are six primary loops with horizontal steam generators and main gate valves. The structure of the horizontal steam generators construction may cause a large primary to secondary leak in case of a break in the cover of the primary collector. An accident where two primary collector covers opened totally and two covers opened partly took place in Rovno, Ukraine January 1982. Primary to secondary leaks are one of the main contributors to the core melt frequency in VVER reactors according to the Loviisa 1 Probabilistic Safety Assessment. The high core damage contribution has set requirements for the development of effective means to cope with all sizes of primary to secondary leaks in the steam generator. A concept for all leak sizes has been developed for Loviisa 1 and 2. The solution includes four main areas which are a new steam generator leakage monitoring system based on nitrogen-16 measurement, an upgraded pressurizer spray system, an increased emergency cooling water reserve and an automated isolation of the defected steam generator.

  4. Spatial Kinetics Calculations of MOX Fueled Core: Variant 22

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-01-11

    This work is part of a Joint US/Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, to compare in future point and spatial kinetics models (on the base of a set of selected accidents) in view of eventual creation of RELAP option with 3D kinetics. The document contains input data and results of model operation of three emergency dynamic processes in the VVER-1000 core: central control rod ejection by pressure drop caused by destroying of the moving mechanism cover; overcooling of the reactor core caused by steam line rupture and non-closure of steam generator stop valve; and the boron dilution of coolant in part of the VVER-1000 core caused by penetration of the distillate slug into the core at start up of non-working loop.

  5. Experiments with the HORUS-II test facility

    Energy Technology Data Exchange (ETDEWEB)

    Alt, S.; Lischke, W. [Univ. for Applied Sciences Zittau/Goerlitz, Zittau (Germany). Dept. of Nuclear Engineering

    1997-12-31

    Within the scope of the German reactor safety research the thermohydraulic computer code ATHLET which was developed for accident analyses of western nuclear power plants is more and more used for the accident analysis of VVER-plants particularly for VVER-440,V-213. The experiments with the HORUS-facilities and the analyses with the ATHLET-code have been realized at the Technical University Zittau/Goerlitz since 1991. The aim of the investigations was to improve and verify the condensation model particularly the correlations for the calculation of the heat transfer coefficients in the ATHLET-code for pure steam and steam-noncondensing gas mixtures in horizontal tubes. About 130 condensation experiments have been performed at the HORUS-II facility. The experiments have been carried out with pure steam as well as with noncondensing gas injections into the steam mass flow. The experimental simulations are characterized as accident simulation tests for SBLOCA for VVER-conditions. The simulation conditions had been adjusted correspondingly to the parameters of a postulated SBLOCA`s fourth phase at the original plant. 4 refs.

  6. Minimization of the fission product waste by using thorium based fuel instead of uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, A. Abdelghafar, E-mail: Agalahom@yahoo.com

    2017-04-01

    This research discusses the neutronic characteristics of VVER-1200 assembly fueled with five different fuel types based on thorium. These types of fuel based on mixing thorium as a fertile material with different fissile materials. The neutronic characteristics of these fuels are investigated by comparing their neutronic characteristics with the conventional uranium dioxide fuel using the MCNPX code. The objective of this study is to reduce the production of long-lived actinides, get rid of plutonium component and to improve the fuel cycle economy while maintaining acceptable values of the neutronic safety parameters such as moderator temperature coefficient, Doppler coefficient and effective delayed neutrons (β). The thorium based fuel has a more negative Doppler coefficient than uranium dioxide fuel. The moderator temperature coefficient (MTC) has been calculated for the different proposed fuels. Also, the fissile inventory ratio has been calculated at different burnup step. The use of Th-232 as a fertile material instead of U-238 in a nuclear fuel is the most promising fuel in VVER-1200 as it is the ideal solution to avoid the production of more plutonium components and long-lived minor actinides. The reactor grade plutonium accumulated in light water reactor with burnup can be recycled by mixing it with Th-232 to fuel the VVER-1200 assembly. The concentrations of Xe-135 and Sm-151 have been investigated, due to their high thermal neutron absorption cross section.

  7. Manufacturing of Ferritic Low-Silicon and Molybdenum Ductile Cast Iron with the Innovative 2PE- 9 Technique

    Directory of Open Access Journals (Sweden)

    Guzik E.

    2014-06-01

    Full Text Available W pracy przedstawiono analizę wyników badań otrzymanych podczas produkcji żeliwa sferoidalnego typu SiMo, z zastosowaniem nowej metody sferoidyzacji metalu w kadzi bębnowej (technika 2PE- 9. Zaprezentowano wyniki badań w zakresie optymalizacji parametrów procesu, takich jak: długości przewodu sferoidyzującego. krytycznej zawartość magnezu, temperatur' zabiegu i temperatury zalewania. Pokazano wpływ temperatur i zabiegu, prędkości przemieszczania przewodu sferoidyzującego (czasu zabiegu sferoidyzowania i masy ciekłego stopu na uzysk magnezu ze sferoidyzatora. Przedstawiono mikrostrukturę, właściwości mechaniczne i koszt wytwarzania terrytycznego żeliwa sferoidalnego SiMo: gatunku EN-GJS-SiMo40-6. zgodnie z najnowszą EN 16124:2011 (E. Wprowadzenie dwóch przewodów elastycznych o średnicy Ø 9 mm; jeden wypełniony mieszaniną FeSi + Mg, a drugi moyfikatorem grafityzującym do zabiegowej kadzi bębnowej, jest nową metodą obróbki pozapiecowej produkcji terrytycznego żeliwa typu SiMo. która może być wykorzystana do produkcji żeliwa sferoidalnego wytapianego w indukcyjnym piecu.

  8. Úzkopásmové filtry pro signály EKG

    OpenAIRE

    Strouhal, Adam

    2009-01-01

    Cílem této práce je odfiltrování nejčastějšího typu rušení v EKG signálech, síťového brumu a driftu. Zabývá se návrhem a realizací lineárních filtrů typu FIR a IIR v prostředí Matlab. Pro realizaci FIR filtrů jsou použity metody vzorkování frekvenční charakteristiky a váhování impulsní charakteristiky, pro IIR filtry metoda návrhu filtru z rozložení nulových bodů a pólů v rovině „z“ a metoda založená na podobnosti s analogovými filtry. The aim of this paper is to filter out the most common...

  9. Smart City objekty a jejich oceňování

    OpenAIRE

    Uheríková, Eliška

    2017-01-01

    Diplomová práce se zaměřuje na specifikování rozdílů budov běžného typu a budov ve smyslu Smart City. Práce obsahuje mimo jiné i historii jednotlivých nově vzniklých kategorií budov, včetně jejich specifikací. Cílem práce je stanovit návrhy na změnu oceňování nákladovým způsobem pro budovy, konkrétně rodinné domy, ve standardu Smart City. Teoretická část obsahuje základní definice a vymezení pojmů z oblasti stavebnictví a typu konstrukcí, krátké shrnutí historického vývoje rodinných domů a me...

  10. Variant 22: Spatially-Dependent: Transient Processes in MOX Fueled Core

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-09-28

    This work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, and to compare in future point and spatial kinetics models (on the base of a set of selected accidents) in view of eventual creation of RELAP option with 3D kinetics. The document contains input data and results of model operation of three emergency dynamic processes in the VVER-1000 core: (1) Central control rod ejection by pressure drop caused by destroying of the moving mechanism cover. (2) Overcooling of the reactor core caused by steam line rupture and non-closure of steam generator stop valve. (3) The boron dilution of coolant in part of the VVER-1000 core caused by penetration of the distillate slug into the core at start up of non-working loop. These accidents have been applied to: (1) Uranium reference core that is the so-called Advanced VVER-1000 core with Zirconium fuel pins claddings and guide tubes. A number of assemblies contained 18 boron BPRs while first year operating. (2) MOX core with about 30% MOX fuel. At a solving it was supposed that MOX-fuel thermophysical characteristics are identical to uranium fuel ones. The calculations were carried out with the help of the program NOSTRA/1/, simulating VVER dynamics that is briefly described in Chapter 1. Chapter 3 contains the description of reference Uranium and MOX cores that are used in calculations. The neutronics calculations of MOX core with about 30% MOX fuel are named ''Variant 2 1''. Chapters 4-6 contain the calculational results of three above mentioned benchmark accidents that compose in a whole the ''Variant 22''.

  11. Zwycięska rewolucja i przegrana modernizacja. Próba parafrazy teorii rewolucji społecznych Thedy Skocpol w aparaturze pojęciowej nie-Marksowskiego materializmu historycznego

    OpenAIRE

    Brzechczyn, Krzysztof

    2013-01-01

    Celem artykułu jest parafraza w aparaturze pojęciowej nie-Marksowskiego materializmu historycznego teorii rewolucji opracowanej przez Thedę Skocpol. W poszczególnych rozdziałach artykułu parafrazowane są pojęcia modernizacji, władzy państwowej, społeczeństwa agrarnej biurokracji i mechanizmu zwycięskiej rewolucji społecznej. Dokonana parafraza pozwala wyróżnić dwa typy społeczeństw agrarnej biurokracji. W ekonomicznej odmianie tego typu społeczeństwa zwycięska rewolucja prowadzi do udanej mod...

  12. The Reverse Task of Discrete Mechatronic Vibration Systems With Negative Stiffness and Capacitance Elements

    Directory of Open Access Journals (Sweden)

    Białas Katarzyna

    2015-06-01

    Full Text Available Korzystajac z wybranych metod rozkładu funkcji charakterystycznej w postaci powolnosci zsyntezowano mechatroniczne dyskretne układy drgajace. W kazdym przypadku układ zbudowano z mechanicznego modelu dyskretnego oraz piezo aktuatora typu „stack” połaczonego z zewnetrznym obwodem elektrycznym LxRxCx. Układy zaprojektowano ze wzgledu na wymagania dynamiczne w postaci biegunów i zer. Zewnetrzny układ elektryczny moze wystepowac w róznych konfiguracjach. W pracy, po bezwymiarowych transformacjach i retransformacjach, zbadano wpływ ujemnych parametrów: sztywnosci w mechanicznym modelu zastepczym oraz pojemnosci w finalnym układzie mechatronicznym LxCx, na charakterystyki rozwazanych układów

  13. Výkon síťového serveru při komunikaci s velkým počtem klientů

    OpenAIRE

    Mašín, Jan

    2008-01-01

    Úkolem je vytvořit simulační systém mezi dvěma počítači typu server-klient. Z klienta generovat provoz pomocí UDP protokolu na server. Generovanému provozu nastavovat různé parametry (velikost paketu, interval zasílání) a vyhodnocovat spotřebu zdrojů na serveru (procesor, paměť). Dosažené výsledky přehledně graficky zpracovat. Byly otestovány dva servery o různé výkonnosti a operační systémy Windows XP, Windows Server 2003, Linux Mandriva 8.0 a Linux Fedora 7, které byly nainstalovány na jedn...

  14. Badanie procesu dehydratacji w hydrożelowych i silikonowo-hydrożelowych soczewkach kontaktowych

    OpenAIRE

    Krysztofiak, Katarzyna

    2016-01-01

    Tematykę dehydratacji miękkich soczewek kontaktowych podjęto ze względu na rosnącą popularność tego typu korekcji. Niestety, mimo znaczącego postępu inżynierii biomateriałów, użytkownicy soczewek wciąż skarżą się na objawy dyskomfortu, takie jak uczucie suchości oczu. Objawy te często wiąże się z nadmierną utratą wody przez materiał soczewek. Celem pracy było badanie dehydratacji soczewek kontaktowych użytkowanych w trybie jednodniowym w kontekście tzw. modelu dyskretnego rozkładu wody w p...

  15. A Comparison of Type A Behaviour Pattern in Cardiovascular High Risk and Normal Adolescents

    OpenAIRE

    Ogińska-Bulik, Nina

    2006-01-01

    Celem podjętych badań było ustalenie czy młodzież z rozpoznanymi czynnikami ryzyka niedokrwiennej choroby serca różni się˛ od młodzieży bez takich czynników w zakresie nasilenia zachowań typu A oraz ustalenie związku między wzorem zachowania A a tymi czynnikami w grupie ryzyka. Badaniami objęto młodzież w wieku 15–18 lat. Część´ pierwsza˛ badań mająca˛na celu wykrycie biologicznych czynników ryzyka choroby niedokrwiennej serca przeprowadzono wśród 350 nastolatków. Część drugą˛ obejmującą pomi...

  16. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  17. Exploratory Study of Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chernobaeva, A.A., Kryukov, A.M., Nikolaev, Y.A., Korolev, Y.N. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)], Sokolov, M.A., Nanstad, R.K. [Oak Ridge National Lab., TN (United States)

    1997-12-31

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVS) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The working group agreed that each side would irradiate, anneal, reirradiate (if feasible), and test two materials of the other; so far, only charpy impact and tensile specimens have been included. Oak Ridge National Laboratory (ornl) conducted such a program (irradiation and annealing) with two weld metals representative of VVER-440 AND VVER-1000 RPVS, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation,annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) program plate 02 and Heavy-Section Steel Irradiation (HSSI) program weld 73w. The results for each material from each laboratory are compared with those from the other laboratory. the ORNL experiments with the VVER welds included irradiation to about 1 x 10 (exp 19) N/SQ CM ({gt}1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 X 10 (exp 19) N/SQ CM ({gt}1 MeV).

  18. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    OpenAIRE

    Ma, Weimin; Yuan, Yidan; Sehgal, Bal Raj

    2016-01-01

    A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential...

  19. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M.; Boehmert, J.; Gilles, R. [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  20. Sensitivity coefficients for the stochastic estimation of the radiation damage to the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, C.M.; Hernandez Valle, S. [Centro de Investigaciones Tecnologicas, Nucleares y Ambientales, La Habana (Cuba). E-mail: calvarez@ctn.isctn.edu.cu; svalle@ctn.isctn.edu.cu

    2000-07-01

    The construction of the sensitivity matrix in the case of the vessel radiation damage estimation by Monte Carlo techniques poses new problems related to the uncertainties of the obtained responses. In the case of deterministic calculations, the sensitivity coefficient obtention is a straightforward procedure based on the perturbation formalism through the calculation of the adjoint fluxes. In the paper an alternative procedure implementation based on the differential operator method is described with the modifications needed to the used HEXANN-EVALU code for the response estimations in the VVER-440 pressure vessel. (author)

  1. Validation of depletion codes for burnup credit evaluation of LWR assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ranta-aho, A. [Technical Research Centre of Finland VTT, POB 1000, 02044-VTT (Finland)

    2006-07-01

    This paper reports the comparison of the CASMO-4E predictions with the radiochemical assay data from assemblies irradiated in Takahama-3 PWR and Fukushima-Daini-2 BWR, and the most recently reported spent fuel data from the VVER-440 assembly irradiated in Novovoronezh 4. Some of the calculations were repeated with the ABURN burnup code, which is a combination of the MCNP4C Monte Carlo code and the ORIGEN2 depletion code. The cross section libraries applied were based on the ENDF/B-VI and the JEF-2.2 data. (authors)

  2. Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

    2016-12-15

    Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

  3. FBIS report. Science and technology: Central Eurasia, April 25, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-25

    ;Partial Contents: Russia: Method for Determining Hardening Diagrams of Individual Structue Components in Multicomponent Systems Developed; Russia: Four-Phase Power Transmission Proposed; Russia: Decontamination Systems in Atomic Power Plant Designs with VVER-640 Reactors; Russia: Synchronization in a Neural Network of Phase Oscillators with Time-Delayed Coupling; Russia: Selection of Prolonging Agent for Eye Drops of Human Recombinant Epidermal Growth Factor; Russia: Double-Glume Mutant of Barley; Synthesis and Study of Cytotoxi Properties of Immunotoxins Based on Bacterial Toxin Phosphol Phospholipase; and Russia: Experience of Organizing the Production of Medical Equipment at Military-Industrial Complex.

  4. Thermal-hydraulics of PGV-4 water volume during damage of the feedwater collector nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S.A.; Titov, V.F. [OKB Gidropress (Russian Federation); Notaros, U.; Lenkei, I. [NPP Paks (Hungary)

    1995-12-31

    A number of VVER-440 plants has experienced the distributing nozzles of feedwater collector being damaged due to corrosion-erosion wearing. Such phenomenon could result in feedwater redistribution within the SG inventory with undesirable consequences. The collector with damaged nozzles has to be replaced but a certain time is needed for the preparatory works. The main objective of the investigation conducted is to assess if the safe operation of SG is possible before collector replacement. It was shown that the nozzle damage as observed did not result in the dangerous disturbances of thermobydraulics as compared with the conditions existing at the initial period of operation. (orig.).

  5. Radiation damage of structural materials

    CERN Document Server

    Koutsky, Jaroslav

    1994-01-01

    Maintaining the integrity of nuclear power plants is critical in the prevention or control of severe accidents. This monograph deals with both basic groups of structural materials used in the design of light-water nuclear reactors, making the primary safety barriers of NPPs. Emphasis is placed on materials used in VVER-type nuclear reactors: Cr-Mo-V and Cr-Ni-Mo-V steel for RPV and Zr-Nb alloys for fuel element cladding. The book is divided into 7 main chapters, with the exception of the opening one and the chapter providing a phenomenological background for the subject of radiation damage. Ch

  6. Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Pintor, S.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)

    2012-07-01

    Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

  7. Design Studies of ``100% Pu'' Mox Lead Test Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-01-11

    In this document the results of neutronics studies of <<100%Pu>> MOX LTA design are presented. The parametric studies of infinite MOX-UOX grids, MOX-UOX core fragments and of VVER-1000 core with 3 MOX LTAs are performed. The neutronics parameters of MOX fueled core have been performed for the chosen design MOX LTA using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.

  8. Experimental study of undeveloped nucleate boiling on the horizontal tube heated by condensing steam

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, A.V.; Remizov, O.V.; Tzyganok, A.A.; Kalyakin, D.S. [State Scientific Center of the Russian Federation, Inst. for Physics and Power Engineering named after A.I. Leypunsky, Obninsk (Russian Federation)

    2011-07-01

    The experimental study of undeveloped nucleate boiling on the horizontal tube heated by condensing steam has been carried out in the Institute for Physics and Power Engineering. The feature of the processes investigated was a presence of natural circulation in primary circuit of the facility. The experiments were carried out at heating steam pressure P{sub s1} = 0.35 MPa. On the base of the results of these experiments the empirical correlations for prediction of heat transfer coefficient was obtained. This correlation can be used for the substantiation of work of VVER steam generator in the condensation mode. (author)

  9. The nature thickness pipe element testing method to validate the application of LBB conception

    Energy Technology Data Exchange (ETDEWEB)

    Vasilchenko, G.S.; Artemyev, V.I.; Merinov, G.N. [and others

    1997-04-01

    To validate the application of leak before break analysis to the VVER-1000 reactor, a procedure for testing a large-scale specimen on electrohydraulic machinery was developed. Steel pipe with a circular weld and stainless cladding inside was manufactured and large-scale longitudinal cross-sections were cut. The remaining parts of the weld after cut out were used to determination standard tensile mechanical properties, critical temperature of brittlness and for manufacture of compact specimens. Experimental mechanical properties of the weld are summarized.

  10. The concepts of leak before break and absolute reliability of NPP equipment and piping

    Energy Technology Data Exchange (ETDEWEB)

    Getman, A.F.; Komarov, O.V.; Sokov, L.M. [and others

    1997-04-01

    This paper describes the absolute reliability (AR) concept for ensuring safe operation of nuclear plant equipment and piping. The AR of a pipeline or component is defined as the level of reliability when the probability of an instantaneous double-ended break is near zero. AR analysis has been applied to Russian RBMK and VVER type reactors. It is proposed that analyses required for application of the leak before break concept should be included in AR implementation. The basic principles, methods, and approaches that provide the basis for implementing the AR concept are described.

  11. Criticality calculation of non-ordinary systems

    Energy Technology Data Exchange (ETDEWEB)

    Kalugin, A. V., E-mail: Kalugin-AV@nrcki.ru; Tebin, V. V. [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    The specific features of calculation of the effective multiplication factor using the Monte Carlo method for weakly coupled and non-asymptotic multiplying systems are discussed. Particular examples are considered and practical recommendations on detection and Monte Carlo calculation of systems typical in numerical substantiation of nuclear safety for VVER fuel management problems are given. In particular, the problems of the choice of parameters for the batch mode and the method for normalization of the neutron batch, as well as finding and interpretation of the eigenvalue spectrum for the integral fission matrix, are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, Juan-Andres, E-mail: lozano@din.upm.e [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid (UPM), Jose G. Abascal, 2, 28006 Madrid (Spain); Jimenez, Javier; Garcia-Herranz, Nuria; Aragones, Jose-Maria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid (UPM), Jose G. Abascal, 2, 28006 Madrid (Spain)

    2010-03-15

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

  13. Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Temesvari, Emese [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

    2017-11-15

    The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.

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

    Science.gov (United States)

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

    2017-09-01

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

  15. Application of the nodal method RTN-0 for the solution of the neutron diffusion equation dependent of time in hexagonal-Z geometry; Aplicacion del metodo nodal RTN-0 para la solucion de la ecuacion de difusion de neutrones dependiente del tiempo en geometria hexagonal-Z

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J.; Alonso V, G. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: jaime.esquivel@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico)

    2015-09-15

    The solution of the neutron diffusion equation either for reactors in steady state or time dependent, is obtained through approximations generated by implementing of nodal methods such as RTN-0 (Raviart-Thomas-Nedelec of zero index), which is used in this study. Since the nodal methods are applied in quadrangular geometries, in this paper a technique in which the hexagonal geometry through the transfinite interpolation of Gordon-Hall becomes the appropriate geometry to make use of the nodal method RTN-0 is presented. As a result, a computer program was developed, whereby is possible to obtain among other results the neutron multiplication effective factor (k{sub eff}), and the distribution of radial and/or axial power. To verify the operation of the code, was applied to three benchmark problems: in the first two reactors VVER and FBR, results k{sub eff} and power distribution are obtained, considering the steady state case of reactor; while the third problem a type VVER is analyzed, in its case dependent of time, which qualitative results are presented on the behavior of the reactor power. (Author)

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

    Directory of Open Access Journals (Sweden)

    Ternovykh Mikhail

    2017-01-01

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

  17. Results of questionnaire for the needs of measured data for the steady-state calculations

    Energy Technology Data Exchange (ETDEWEB)

    Yrjoelae, V. [VTT Energy, Espoo (Finland)

    1995-12-31

    In the First International Seminar on the Modelling of Horizontal Steam Generators arranged in March 1991 was agreed to arrange a common calculational exercise to calculate the secondary side flow conditions during normal plant operation. OKB Gidropress of Russia supplied the experimental results for the exercise. They included some measured data of the local velocities and void fractions for the steam generators of the VVER-440 and VVER-1000 type reactors. The results of the common calculational exercise presented in the Second International Seminar in September 1992 were still mainly preliminary and it was felt necessary to continue these efforts. It was concluded that the given experimental results were not sufficient for a real code assessment - still too many quantities have to be guessed. It was pointed out that it is advisable to define a minimum set of necessary data. For this reason it was decided that VTT should made a query among the participants of the seminar, where they can give their opinion of the essential data. In this presentation the results of the questionnaire are given.

  18. PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 2: User information

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J. [Argonne National Lab., IL (United States). Reactor Engineering Div.

    1997-06-01

    A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. PACER has been developed for FORTRAN 77 and earlier versions of FORTRAN. The code has been successfully compiled and executed on SUN SPARC and Hewlett-Packard HP-735 workstations provided that appropriate compiler options are specified. The code incorporates both capabilities built around a hardwired default generic VVER-440 Model V230 design as well as fairly general user-defined input. However, array dimensions are hardwired and must be changed by modifying the source code if the number of compartments/cells differs from the default number of nine. Detailed input instructions are provided as well as a description of outputs. Input files and selected output are presented for two sample problems run on both HP-735 and SUN SPARC workstations.

  19. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-06-16

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10{sup 19} n/cm{sup 2} (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10{sup 19} n/cm{sup 2} (>l MeV). In both cases, irradiations were conducted at {approximately}290 C and annealing treatments were conducted

  20. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    Energy Technology Data Exchange (ETDEWEB)

    Foehl, J.; Weissenberg, T. [Materialpruefungsanstalt, Univ. Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT) (Spain); Ernestova, M.; Zamboch, M. [Nuclear Research Inst. (NRI) (Czech Republic); Seifert, H.P.; Ritter, S. [Paul Scherrer Inst. (PSI) (Switzerland); Roth, A.; Devrient, B. [Framatome ANP GmbH (F ANP) (Germany); Ehrnsten, U. [Technical Research Centre of Finland (VTT) (Finland)

    2004-07-01

    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  1. Immunoterapia podjęzykowa

    Directory of Open Access Journals (Sweden)

    Anna Grad

    2009-09-01

    Full Text Available Immunoterapia swoista (SIT stanowi jedyne, obok eliminacji alergenu z otoczenia chorego, leczenie przyczynowe chorób alergicznych. Polega na podawaniu choremu wzrastających dawek szczepionki alergenowej drogą iniekcyjną (SCIT lub rzadziej stosowaną drogą nieiniekcyjną. Do immunoterapii miejscowej zaliczamy: immunoterapię podjęzykową typu spit i typu swallow, immunoterapię doustną, donosową, dooskrzelową i dospojówkową. Pierwsze próby stosowania immunoterapii doustnej sięgają roku 1900. Przez kolejne ponad sto lat ta forma leczenia cieszyła się zmiennym zainteresowaniem. W latach 90. XX wieku wyraźnie wzrosło zainteresowanie nieiniekcyjnymi metodami immunoterapii swoistej. Badania ostatnich lat potwierdzają, że immunoterapia podjęzykowa może być alternatywą dla immunoterapii iniekcyjnej. Warunkiem koniecznym jest stosowanie wysokich dawek alergenu. SLIT jest bezpiecznym i akceptowanym przez chorego sposobem leczenia. Większość obserwowanych objawów ubocznych wiąże się z bezpośrednim narażeniem błony śluzowej na działanie szczepionki alergenowej. Objawy miejscowe są zwykle krótkotrwałe, ustępują samoistnie i nie wymagają leczenia. Szczepionki używane do immunoterapii swoistej powinny być standaryzowane za pomocą testów biologicznych. Dzięki temu kolejne serie szczepionek są porównywalne z wzorcem biologicznym pod względem zawartości alergenu i siły działania. Zaleca się stosowanie w szczepionce maksymalnie 4 alergenów oraz niełączenie w niej alergenów sezonowych i całorocznych. Immunoterapię swoistą można prowadzić sezonowo i całorocznie, schemat całoroczny umożliwia podanie wyższych dawek szczepionki alergenowej, aw związku z tym osiągnięcie lepszej skuteczności klinicznej. Optymalny okres prowadzenia immunoterapii swoistej nie został ostatecznie określony, zaleca się prowadzenie SIT nie krócej niż 3 do 5 lat.

  2. Effect of Magnesium Addition on Properties of Al-Based Composite Reinforced with Fine NiO Particles

    Directory of Open Access Journals (Sweden)

    Zygmunt-Kiper M.

    2014-06-01

    Full Text Available Kompozyt Al(Mg-NiO wytworzono metodą mechanicznej syntezy stosując mielenie składników proszkowych i mechaniczną konsolidację uzyskanego proszku kompozytowego w procesie prasowania próżniowego i wyciskania „na gorąco”. Uzyskano jednorodny materiał charakteryzujący się dużym rozdrobnieniem składników strukturalnych. Próby ściskania w temperaturze 293 K - 773 K wykazały wysokie własności mechaniczne kompozytu. Wyżarzanie próbek w 823 K / 6 godz. spowodowało nieznaczne obniżenie wartości naprężenia uplastyczniającego w zakresie 573 K - 773 K, jednakże w znacznie mniejszym stopniu niż w porównywanym przypadku kompozytu nie zawierającego dodatku magnezu (Al-NiO. który opisano we wcześniejszej pracy. Obserwacje struktury wyjściowych próbek kompozytowych i próbek wyżarzonych w 823 K / 6 godz. wykazały zmiany strukturalne wywołane reakcją chemiczną między osnową kompozytu (Al-Mg a dyspersyjnymi cząstkami zbrojenia (NiO, której skutkiem jest utworzenie silnie dyspersyjnych wydzieleń tlenków typu spinelu, oraz submikronowych ziarn typu Al3Ni. W porównaniu z kompozytem Al-NiO, dodatek magnezu powoduje zwiększenie szybkości reakcji chemicznej, która przejawia się utworzeniem ziarn fazy międzymetalicznej Al3Ni zarówno w materiale wyjściowym - wyciskanym „na gorąco” - jak również w próbkach wyżarzonych w 823 K / 6 godz.

  3. Evaluación de la toxicidad aguda oral y de la actividad antimicrobiana de una mezcla de aceite de hígado de tiburones de Cuba Assessment of the oral acute toxicity and the antimicrobial activity of an oily mixture from shark's liver of Cuba

    Directory of Open Access Journals (Sweden)

    Caridad Margarita García Peña

    2010-09-01

    Full Text Available Se evaluó la toxicidad aguda oral y la actividad antimicrobiana de una mezcla de aceites de hígado de tiburón, de las especies Rhincodon typu (tiburón ballena y Galeocerdo cuvier (tiburón tigre, que habitan en zonas aledañas a las costas del litoral norte occidental de Cuba, para su posterior uso farmacéutico, debido a que presenta un alto contenido de vitaminas y de ácidos grasos, que le confieren actividad antioxidante y antiinflamatoria. El estudio de la toxicidad aguda oral demostró que la mezcla de aceites de hígado de tiburones, no provocó alteraciones macroscópicas en los órganos extraídos, ni síntomas tóxicos severos, ni mortalidad de ninguno de los animales empleados en el estudio a la dosis de 20 mL/kg. Los resultados del estudio de la actividad antimicrobiana demostraron una ligera actividad bacteriostática frente a K. pneumoniae; además una actividad antifúngica frente a Microsporum canis; y resistencia frente a C. albicans y T. mentagrophytes a las concentraciones evaluadas.The total acute toxicity and the antimicrobial activity of an oil mixtures from shark liver of Rhicodon typu (whale-shark and Galeocerdo cuvier (tigger-shark was assessed in species leaving in the adjacent costs of Cuban northern coastal for its subsequent pharmaceutical use due to its high content of vitamins and fatty acids and its antioxidant and anti-inflammatory activity. Study of oral acute toxicity demonstrated that oil mixture of shark liver hasn't macroscopic alterations in removed organs, severe toxic symptoms and on mortality of any animals used in study at 20 mL/kg dose. Study results of antimicrobial activity showed a slight bacteriostatic activity against K. pneumoniae and an antifungal activity against Microsporum canis, and a resistance against C. albicans and T. mentagrophytes at assessed concentrations.

  4. Development of an improved methodology for the determination of the neutron load of the pressure vessel steel of WWER-1000 type reactors. Final report; Entwicklung einer fortgeschrittenen Methodik zur Bestimmung der Neutronenbelastung des Druckbehaeltermaterials vom Reaktor des Typs WWER-1000. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Barz, H.U.; Boehmer, B.; Konheiser, J.; Stephan, I.

    1998-05-01

    The investigations, reported on here, are aimed at the theoretical and experimental determination of the neutron exposure of the VVER-1000 reactors Balakovo-3 and Rovno-3. The overall approach, partially developed and at least improved in the frame of this project, comprises the pure calculation part, the gamma spectrometric analysis of neutron activation detectors and the comparison of experimental and theoretical results using the spectrum adjustment procedure. The approach is not restricted to neutron embrittlement but can be applied to neutron fluence problems in general. (orig./GL) [Deutsch] In diesem Projekt wurden fuer die WWER-1000 Reaktoren Balakovo-3 und Rovno-3 die Parameter der Neutronenbelastung experimentell und theoretisch bestimmt. Der vorliegende Bericht beschreibt das methodische Vorgehen, welches aus dem reinen Berechnungsteil, der gammaspektrometrischen Analyse der Ativierungsdetektoren und dem Vergleich der gemessenen und berechneten Werte einschliesslich der Spektrumsjustierung besteht. Dieses Instrumentarium, welches allgemein bei der Bestimmung der Neutronenfluenz anwendbar ist, wurde im Projektzeitraum weiter verbessert. (orig./GL)

  5. The prospect of nuclear energy in Türkiye especially after Fukushima accident

    Science.gov (United States)

    Şahin, Sümer

    2014-09-01

    Türkiye considers since mid-50's to use nuclear electricity, but Government and bureaucracy have continuously postponed reactor construction. However, since 2010 the case has gained a real shape. Official agreement has been signed for the construction of 4 units of Russian VVER type reactors with installed power of 4×1200 MWel. It is expected that they will begin to deliver electricity early 20's. Further negotiations are being conducted with Japanese Mitsubashi and French AREVA. The target is to have nuclear electricity by 2023 at the 100th anniversary of Turkish Republic. Turkish Nuclear Energy Strategy aims; • Decrease country's dependency on foreign suppliers of energy sources • Provide fuel supply mix diversification • Utilization of environmentally friendly energy production technologies Possess advanced and prestigious power generation technologies.

  6. In-vessel melt retention as a severe accident management strategy for the Loviisa Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kymaelaeinen, O.; Tuomisto, H. [IVO International Ltd., Vantaa (Finland); Theofanous, T.G. [Univ. of California, Santa Barbara, CA (United States)

    1997-02-01

    The concept of lower head coolability and in-vessel retention of corium has been approved as a basic element of the severe accident management strategy for IVO`s Loviisa Plant (VVER-440) in Finland. The selected approach takes advantage of the unique features of the plant such as low power density, reactor pressure vessel without penetrations at the bottom and ice-condenser containment which ensures flooded cavity in all risk significant sequences. The thermal analyses, which are supported by experimental program, demonstrate that in Loviisa the molten corium on the lower head of the reactor vessel is coolable externally with wide margins. This paper summarizes the approach and the plant modifications being implemented. During the approval process some technical concerns were raised, particularly with regard to thermal loadings caused by contact of cool cavity water and hot corium with the reactor vessel. Resolution of these concerns is also discussed.

  7. Monitoring and diagnostics systems for nuclear power plant operating regimes

    Energy Technology Data Exchange (ETDEWEB)

    Abagyan, A.A.; Dmitriev, V.M.; Klebanov, L.A.; Kroshilin, A.E.; Larin, E.P.; Morozov, S.K.

    1988-05-01

    The development of new monitoring and diagnostics systems for Soviet reactors is discussed. An experimental test station is described where industrial operation of new experimental systems can be conducted for purposes of bringing their performance to the level of standard Soviet systems for monitoring reactor operation regimes and equipment resources. The requirements and parameters of the systems are described on a unit-by-unit basis, including the sensor reading monitoring unit, the vibroacoustic monitoring unit, the noise monitoring unit, the accident regime identification unit, and the nonstationary regime monitoring unit. Computer hardware and software requirements are discussed. The results of calculational and experimental research on two complex nonstationary regimes of reactor operation are given. The accident regimes identification unit for the VVER-1000 is analyzed in detail.

  8. The hydrological impact assessment in the decision support of nuclear emergency response.

    Science.gov (United States)

    Vamanu, Dan V; Slavnicu, Dan S; Gheorghiu, Dorina; Acasandrei, Valentin T; Slavnicu, Elena

    2010-07-01

    The paper presents several aspects believed to be relevant for the integration in the decision support systems for the management of radiological emergencies, of assessment tools addressing surface water contamination. Three exemplary cases are discussed in the context-the CONVEX 2005 international alert exercise, AXIOPOLIS 09, a national drill targeting a CANDU reactor at Cernavoda nuclear power plant in Romania, and Oltenia 07-a nation-wide drill around a scenario, involving trans-border effects of a virtual accident at a VVER reactor at Kozloduy, Bulgaria. The capability of different analytic tools were tested, including public deliverables like real-time, online decision support system's HDM module and model-based computerised system for management support to identify optimal remedial strategies for restoring radionuclide-contaminated aquatic ecosystems and drainage areas, as well as research-grade, home-made facilities, in order to identify and sort out merits and issues of interest in steering their effective utilisation.

  9. Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

    Energy Technology Data Exchange (ETDEWEB)

    D' Hondt, P. [SCK.CEN, Mol (Belgium); Gehin, J. [ORNL, Oak Ridge, TN (United States); Na, B.C.; Sartori, E. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency, 92 - Issy les Moulineaux (France); Wiesenack, W. [Organisation for Economic Co-Operation and Development/HRP, Halden (Norway)

    2001-07-01

    One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

  10. Uncertainties of the neutronic calculations at core level determined by the KARATE code system and the KIKO3D code

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

    2013-09-15

    In this paper the uncertainties of the neutronic calculations at core level - originating from the uncertainties of the basic nuclear data - are presented. The investigations have been made for a VVER-1000 core (Kozloduy-6) defined in the frame of the OECD NEA UAM benchmark. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown. After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP (Hot Zero Power) state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated. In both cases, we will see that the most important quantities - at core level and at HZP state - have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. (orig.)

  11. The transport of fuel assemblies. New containers for transport the used nuclear material in Juzbado factory; Como se transportan los elementos combustibles? Nuevos contenedores para transporte de material fisionable utilizados en la fabrica de Juzbado

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    Juzbado Manufacturing Facility is designed to be versatile and flexible. It is manufactured different kind of fuel assemblies PWR, BWR and VVER, beginning by the uranium oxide coming from the conversion facilities. The transport of these products (radioactive material fissile) requires the availability of different kind of packages; our models variety is similar to the big manufacturers. It is required a depth knowledge of the licensing process, approvals, manufacturing and handling instruction to be confident. Moreover, the recently changes on the Transport Regulations and the demands for the approval by the Competent Authorities have required the renovation of most of the package designs for the transport of radioactive material fissile worldwide. ENUSA assumed time ago this renovation and it is nowadays in the pick moment of this process. If we also consider the complexity on the management of multimodal international transportations, the Logistic task for the transport of nuclear material associated to the Juzbado factory results in a real changeling area. (Author)

  12. Measurement of Neutron Field Characteristics at Nuclear-Physics Instalations for Personal Radiation Monitoring

    CERN Document Server

    Alekseev, A G; Britvich, G I; Kosyanenko, E V; Pikalov, V A; Gomonov, I P

    2003-01-01

    n this work the observed data of neutron spectra on Rostov NEP, Kursk NEP and Smolensk NEP and on the reactor IRT MIPHI are submitted. For measurement of neutron spectra two types of spectrometer were used: SHANS (IHEP design ) and SDN-MS01 (FEI design). The comparison of the data measurements per-formed by those spectrometers above one-type cells on the reactor RBMK is submitted. On the basis of the 1-st horizontal experimental channel HEC-1 of the IRT reactor 4 reference fields of neutrons are investigated. It is shown, that spectra of neutrons of reference fields can be used for imitation of neutron spectra for conditions of NEP with VVER and RBMK type reactors.

  13. Modeling of Flow in Nuclear Reactor Fuel Cell Outlet

    Directory of Open Access Journals (Sweden)

    František URBAN

    2010-12-01

    Full Text Available Safe and effective load of nuclear reactor fuel cells demands qualitative and quantitative analysis of relations between coolant temperature in fuel cell outlet temperature measured by thermocouple and middle temperature of coolant in thermocouple plane position. In laboratory at Insitute of thermal power engineering of the Slovak University of Technology in Bratislava was installed an experimental physical fuel cell model of VVER 440 nuclear power plant with V 213 nuclear reactors. Objective of measurements on physical model was temperature and velocity profiles analysis in the fuel cell outlet. In this paper the measured temperature and velocity profiles are compared with the results of CFD simulation of fuel cell physical model coolant flow.

  14. Construction prospects of new power units at Khmelnitskij NPP site

    Energy Technology Data Exchange (ETDEWEB)

    Zenyuk, Denys [NNEGC ' Energoatom' , 01032 Vetrova, 3, Kiev (Ukraine)

    2008-07-01

    According to the Energy Strategy of Ukraine for a period up to 2030 it is planned to put into operation power units 3 and 4 of Khmelnitskij NPP by year 2016. In this work considerations are presented on the possible options while selecting reactor unit type for Khmelnitskij NPP power units 3 and 4, which is the main determinant of the cost, construction and commissioning time, and utilization of the existent civil structures. To optimize Khmelnitskij-3 and 4 construction, a survey of the data has been conducted with regard to the possibility of construction of new power units of PWR/VVER type at Khmelnitskij NPP site. The multivariable analysis has been performed based on the projects technical and cost data, construction time and conditions, as well as their compliance with the IAEA and EUR safety requirements for new power units. (author)

  15. PACTEL: Experiments on the behaviour of the new horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

    1995-12-31

    Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

  16. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, RJ

    2001-02-02

    The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

  17. Numerical investigation of mass transfer in the flow path of the experimental model of the PGV-1500 steam generator's steam receiving section with two steam nozzles

    Science.gov (United States)

    Golibrodo, L. A.; Krutikov, A. A.; Nadinskii, Yu. N.; Nikolaeva, A. V.; Skibin, A. P.; Sotskov, V. V.

    2014-10-01

    The hydrodynamics of working medium in the steam volume model implemented in the experimental setup constructed at the Leipunskii Institute for Physics and Power Engineering was simulated for verifying the procedure of calculating the velocity field in the steam space of steam generators used as part of the reactor plants constructed on the basis of water-cooled water-moderated power-generating reactors (VVER). The numerical calculation was implemented in the environment of the STAR-CCM+ software system with its cross verification in the STAR-CD and ANSYS CFX software systems. The performed numerical investigation served as a basis for substantiating the selection of the computation code and parameters for constructing the computer model of the steam receiving device of the PGV-1500 steam generator experimental model, such as the quantization scheme, turbulence model, and mesh model.

  18. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  19. Establishing an effective plant maintenance strategy requires a consortium of information and coordinated resources; Para establecer un programa de mantenimiento eficaz, es necesario conocer desde el inicio el estado de los equipos

    Energy Technology Data Exchange (ETDEWEB)

    Kozsky, T. A.

    2005-07-01

    Establishing an effective plant maintenance strategy requires a consortium of information and coordinated resources. A fundamental element of achieving such a program is to have a comprehensive knowledge base of equipment health derived from operational data, monitoring and diagnostic data, design base data, and others. This paper presents a summary of the elements of a maintenance strategy and focuses on the Westinghouse information integration scheme, called ALLY, that combines equipment health information, performs automated expert system reasoning, organizes and prioritizes equipment health diagnostics, and helps to identify maintenance actions. ALLY has been applied to a VVER nuclear plant located in the Czech Republic, a RBMK nuclear plant located in Rusia and a PWR nuclear plant located in China. A short discription of these applications is also presented. (Author)

  20. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  1. Validation of lattice code 'EXCEL' with TIC experiments on uniform and regularly perturbed lattices

    Energy Technology Data Exchange (ETDEWEB)

    Ramakrishna, A., E-mail: anantatmula.ramakrishna@gmail.co [Atomic Energy Regulatory Board, Niyamak Bhavan, Anushaktinagar, Mumbai 400 094 (India); Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Jain, R.P. [IIT Bombay, Mumbai (India)

    2010-12-15

    Temporary International Collective (TIC) was established in 1972 by an agreement among seven countries, namely, Bulgaria, Czechoslovakia, Germany, Hungary, Poland, Romania and Union of Soviet Socialist Republics. The main objective of TIC was to provide the experimental data for the reactor physics analysis of water cooled and water moderated power reactors (WWER). Extensive experimental work for different core configurations was carried out by TIC countries to investigate the physics behaviour of WWER lattices and the results were published in TIC volumes. Two VVER-1000 MWe reactors are currently in an advanced stage of construction and due for commissioning in Kudankulam, Tamil Nadu, India. Indigenous development of in-core fuel management computer codes for the analysis of hexagonal lattice cores is also in an advanced stage to address various design, operation and safety issues of VVER type cores. The validation of the above TIC lattice experiments will help in the identification of deficiencies in reactor physics design computational codes and the associated nuclear data libraries. In this paper, TIC experiments on uniform and regularly perturbed lattices have been analyzed as part of the validation of indigenous computer codes, EXCEL, TRIHEX-FA and HEXPIN developed at Light Water Reactors Physics Section, B.A.R.C. Neutron-nuclear multi-group cross-section libraries in WIMS/D format in 69/172 energy groups have been released by IAEA at the conclusion of WIMS library update project (WLUP). In the present study we have used libraries based on ENDF/B-6, ENDF/B-7, JEFF3.1 and JENDL3.2 evaluated nuclear datasets. The results of the theoretical analyses bring out the performance of the code system and various cross-section libraries.

  2. Radioactive waste management and plutonium recovery within the context of the development of nuclear energy in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Kushnikov, V. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation)

    1996-05-01

    The Russian strategy for radioactive waste and plutonium management is based on the concept of the closed fuel cycle that has been adopted in Russia, and, to a great degree, falls under the jurisdiction of the existing Russian nuclear energy structures. From its very beginning, Russian atomic energy policy was based on finding the most effective method of developing the new fuel direction with the maximum possible utilization of the energy potential from the fission of heavy atoms and the achievement of fuel self-sufficiency through the recycling of secondary fuel. Although there can be no doubt about the importance of economic considerations (for the future), concerns for the safety of the environment are currently of the utmost importance. In this context, spent NPP fuel can be viewed as a waste to be buried only if there is persuasive evidence that such an approach is both economically and environmentally sound. The production of I GW of energy per year is accompanied by the accumulation of up to 800-1000 kg of highly radioactive fission products and approximately 250 kg of plutonium. Currently, spent fuel from the VVER 100 and the RBNK reactors contains approximately 25 tons of plutonium. There is an additional 30 tons of fuel-grade plutonium in the form of purified oxide, separated from spent fuels used in VVER440 reactors and other power production facilities, as well as approximately 100 tons of weapons-grade plutonium from dismantled warheads. The spent fuel accumulates significant amounts of small actinoids - neptunium americium, and curium. Science and technology have not yet found technical solutions for safe and secure burial of non-reprocessed spent fuel with such a broad range of products, which are typically highly radioactive and will continue to pose a threat for hundreds of thousands of years.

  3. The Activation Detector Activity Calculations Using the Effective Source Method and Measurement

    Science.gov (United States)

    Smutný, Vladimir; Konečná, Alena; Sprinzl, Daniel; Klupák, Vít; Vinš, Miroslav

    2017-09-01

    In the paper the application of effective source to the solution of activation detector activities in the reactor pressure vessel cavity of the VVER-1000 reactor is presented. The effective source method applies the Boltzmann transport operator to time integrated source data to obtain detector activities. Weighting the source data by time dependent depletion of the detector activity, the result of the calculation is the detector activity. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the `effective source'. The effective source method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. Detailed description of the effective source method is presented in previous works. First, there were performed neutron-physical calculations of few real VVER-1000 cycles using MOBY-DICK macrocode. Second, there follows 3-D transport calculation using the deterministic code TORT and the cross section library BUGLE-B7 and obtained results are presented. These calculation results of activation detector activities in the reactor cavity are compared with relevant activation detectors results of the ex-vessel measurement. The comparison between calculation and measurement of activation detectors activity in the reactor cavity is necessary to the calculation quality verifying for further fast neutron fluence onto the reactor pressure vessel credible calculation. The activation detectors positions are evident from Figs 1, 2, 3.

  4. RELAP5 simulation of surge line break accident using combined and best estimate plus uncertainty approaches

    Energy Technology Data Exchange (ETDEWEB)

    Kristof, Marian [AiNS, Na hlinach 51, 917 01 Trnava (Slovakia); Department of Nuclear Physics and Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia)], E-mail: marian.kristof@ains.sk; Kliment, Tomas [VUJE a.s., Okruzna 5, 918 64 Trnava (Slovakia); Petruzzi, Alessandro [Nuclear Research Group of San Piero a Grado, University of Pisa (Italy); Lipka, Jozef [Department of Nuclear Physics and Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia)

    2009-11-15

    Licensing calculations in a majority of countries worldwide still rely on the application of combined approach using best estimate computer code without evaluation of the code models uncertainty and conservative assumptions on initial and boundary, availability of systems and components and additional conservative assumptions. However best estimate plus uncertainty (BEPU) approach representing the state-of-the-art in the area of safety analysis has a clear potential to replace currently used combined approach. There are several applications of BEPU approach in the area of licensing calculations, but some questions are discussed, namely from the regulatory point of view. In order to find a proper solution to these questions and to support the BEPU approach to become a standard approach for licensing calculations, a broad comparison of both approaches for various transients is necessary. Results of one of such comparisons on the example of the VVER-440/213 NPP pressurizer surge line break event are described in this paper. A Kv-scaled simulation based on PH4-SLB experiment from PMK-2 integral test facility applying its volume and power scaling factor is performed for qualitative assessment of the RELAP5 computer code calculation using the VVER-440/213 plant model. Existing hardware differences are identified and explained. The CIAU method is adopted for performing the uncertainty evaluation. Results using combined and BEPU approaches are in agreement with the experimental values in PMK-2 facility. Only minimal difference between combined and BEPU approached has been observed in the evaluation of the safety margins for the peak cladding temperature. Benefits of the CIAU uncertainty method are highlighted.

  5. Air source heat pump

    OpenAIRE

    Krejsa, Petr

    2016-01-01

    Diplomová práce se zabývá popisem technologie tepelných čerpadel (TČ) a jejich využitím pro bytové domy. Teoretická část obsahuje úvod, představení technologie tepelných čerpadel, zejména kompresorového tepelného čerpadla (KTČ). Dále je v této části pojednáno o tepelných čerpadlech pro bytové domy. Hlavní náplní práce je návrh kompaktního tepelného čerpadla typu vzduch-voda pro společnost AISECO spol. s.r.o. V této části jsou navrženy jednotlivé komponenty chladivového oběhu společně s návrhe...

  6. Sběrnice používané v oblasti programovatelných automatů

    OpenAIRE

    Drábek, Jakub

    2008-01-01

    Bakalářská práce je rešeršního typu. Jedná se o shrnutí informací o průmyslových sběrnicích používaných v oblasti programovatelných automatů, které se využívají v laboratoři programovatelných automatů Ústavu automatizace a informatiky na Fakultě strojního inženýrství. Popisovanými průmyslovými sběrnicemi jsou PROFIBUS, PROFINET, AS-Interface, Interbus a CAN. Každá kapitola obsahuje detailní informace o určité sběrnici. Na závěr je uvedeno shrnutí popisovaných sběrnic a jejich srovnání. Thi...

  7. Existuje recept na udržitelný způsob života?

    Directory of Open Access Journals (Sweden)

    Lukáš Hrubý

    2011-12-01

    Full Text Available Udržitelný způsob života – v dnešní době často skloňované téma, obvykle ve spojení s trvale udržitelným rozvojem (jsou dokonce tací, kteří tyto dva pojmy zaměňují. Zatímco v západních zemích se dnes dává mnoho úsilí i peněz do trvale udržitelného způsobu života, státy typu Čína si „vesele jedou“ na co nejvyšší HDP a vysoké zisky, „budoucí generace – nebudoucí generace“. Okolo takovéhoto tématu již jistě existuje velké množství různých teorií, názorů a na dané téma výzkumných prací. Také já se pokusím do této „mísy udržitelnosti“ přispět

  8. The role of vitamin D in the development of autoimmune diseases.

    Science.gov (United States)

    Lisowska, Katarzyna A; Bryl, Ewa

    2017-08-28

    Witamina D, poza istotną rolą w utrzymaniu homeostazy wapnia i metabolizmie kostnym, odgrywa ważną rolę w funkcjonowaniu układu odpornościowego. Niedobór witaminy D wiąże się z wieloma niekorzystnymi dla zdrowia skutkami, włączając w to m.in. osłabienie odporności, czego skutkiem jest zwiększona podatność na zakażenia wirusowe, bakteryjne oraz grzybicze. W artykule opisano podstawy metabolizmu witaminy D oraz jej rolę fizjologiczną, ze szczególnym uwzględnieniem wpływu na komórki układu odpornościowego. Ze względu na jej istotną rolę w regulacji odpowiedzi zapalnej oraz wytwarzaniu cytokin zwraca się uwagę na jej rolę w rozwoju chorób o podłożu autoimmunologicznym, takich jak cukrzyca typu 1, toczeń rumieniowaty, reumatoidalne zapalenie stawów, stwardnienie rozsiane, nieswoiste zapalenia jelit, łuszczyca, bielactwo, czy twardzina, w których witamina D ma potencjalne szerokie zastosowanie zarówno w prewencji, jak i wspomaganiu działań terapeutycznych.

  9. The role of vitamin D in the development of autoimmune diseases

    Directory of Open Access Journals (Sweden)

    Katarzyna A. Lisowska

    2017-08-01

    Full Text Available Witamina D, poza istotną rolą w utrzymaniu homeostazy wapnia i metabolizmie kostnym, odgrywa ważną rolę w funkcjonowaniu układu odpornościowego. Niedobór witaminy D wiąże się z wieloma niekorzystnymi dla zdrowia skutkami, włączając w to m.in. osłabienie odporności, czego skutkiem jest zwiększona podatność na zakażenia wirusowe, bakteryjne oraz grzybicze. W artykule opisano podstawy metabolizmu witaminy D oraz jej rolę fizjologiczną, ze szczególnym uwzględnieniem wpływu na komórki układu odpornościowego. Ze względu na jej istotną rolę w regulacji odpowiedzi zapalnej oraz wytwarzaniu cytokin zwraca się uwagę na jej rolę w rozwoju chorób o podłożu autoimmunologicznym, takich jak cukrzyca typu 1, toczeń rumieniowaty, reumatoidalne zapalenie stawów, stwardnienie rozsiane, nieswoiste zapalenia jelit, łuszczyca, bielactwo, czy twardzina, w których witamina D ma potencjalne szerokie zastosowanie zarówno w prewencji, jak i wspomaganiu działań terapeutycznych.

  10. Bezpečnost IoT

    OpenAIRE

    Hanzlíček, Martin

    2016-01-01

    Práca sa zaoberá problematikou zabezpečenia platformy Android Wear. Rieši časovú náročnosť výpočtov s veľkými číslami, modulárnou aritmetikou a kryptografiou na tejto platforme. Ďalej rieši komunikáciu s mobilným zariadením a autentizačný protokol typu klient-server. Výstupom práce budú tri aplikácie. Jedna bude určená pre Android Wear, ktorá otestuje časovú náročnosť daných výpočtov, komunikáciu s mobilným zariadením a tak tiež bude fungovať ako vydavateľ registrácie a overenia užívateľovi. ...

  11. Výměníky tepla Sodík - Oxid uhličitý pro JE se sodíkem chlazeným rychlým reaktorem (SFR)

    OpenAIRE

    Foral, Štěpán

    2011-01-01

    Tato diplomová práce se zabývá návrhem výměníku tepla sodík-oxid uhličitý. V první části je provedeno porovnání trubkových výměníků tepla s tepelnými výměníky typu PCHE. Dále byl vybrán trubkový výměník tepla s vnitřně žebrovanými trubkami jako základní koncepce. Pro tuto koncepci tepelného výměníku byla provedena optimalizace konstrukčních a provozních parametrů na základě tepelných a hydraulických výpočtů. Dále byly provedeny pevnostní výpočty pro zajištění bezpečného provozu tepelného výmě...

  12. Thermodynamic Description Of Ternary Fe-B-X Systems. Part 2: Fe-B-Ni

    Directory of Open Access Journals (Sweden)

    Miettinen J.

    2014-06-01

    Full Text Available Przedstawiono termodynamiczny opis trójskładnikowego układu Fe-B-Ni w kontekście nowej bazy danych dla układów Fe-B-X (X = Cr, Ni, Mn, Si, Ti, V, C. Parametry termodynamiczne dwuskładnikowych stopów Fe-B. Fe-Ni i B-Ni zostały są zaczerpnięte z wcześniejszych opracowań, przy tym opis B-Ni został nieznacznie zmodyfikowany. Parametry dla układu Fe-B-Ni zostały zoptymalizowane w tej pracy w oparciu o eksperymentalne równowagi fazowe i dane termodynamiczne zaczerpnięte z literatury. Roztwory stałe w układzie Fe-B-Ni opisano przy użyciu modelu roztworu substytucyjnego, a borki traktowane są jako fazy stechiometryczne lub półstechiometryczne typu (A.BpCq opisane przy użyciu modelu dwu podsieci.

  13. História genézy cla v medzinárodnom obchode

    OpenAIRE

    Ladislav Balko

    2007-01-01

    Dejiny ciel sú nerozlučne spojené s dejinami obchodu, ktorý bol úzko spojený s vývojom nového typu výrobných a vlastníckych vzťahov. S inštitútom cla sa stretávame už v starovekom Egypte, Grécku a Ríme. Clo v dnešnom ponímaní možno zaznamenať najskôr vo Veľkej Británii počas anglickej buržoáznej revolúcie v polovici 17. storočia. V ostatných európskych krajinách nastúpil vývoj vedúci k zavedeniu jednotných ciel približne od druhej polovice 18. storočia do polovice 19. storočia. Postupným vývo...

  14. Bird conservation on electric-power lines in Hungary: Nest boxes for saker falcon and avian protection against electrocutions. Projects' report

    Directory of Open Access Journals (Sweden)

    Fidlóczky József

    2014-12-01

    Full Text Available Ochrana vtáctva na elektrickom vedení má v Madarsku 40-rocnú tradíciu. Zacala s programom na ochranu sokola rároha. Pociatocné maloplošné aktivity sa znacne rozšírili v rámci projektov LIFE. Pocas prvého projektu bolo na stlpy vysokého napätia umiestnených 301 hliníkových hniezdnych búdok nového typu pre rároha. V roku 201 3 hniezdilo v týchto búdkach takmer 70 % párov sokola rároha. Odhaduje sa, že rocne zahynulo na stlpoch elektrického vedenia v Madarsku 1 00 000 jedincov vtákov. V rámci dvoch projektov LIFE bolo 1 4 300 stlpov ošetrených tak, že sa stali bezpecnými pre vtáctvo. Pocas druhého projektu sa na stlpy nainštaluje takmer 800 nových, pre vtáky bezpecných konzol, vdaka comu ocakávame kompletné odstránenie rizika úrazov vtáctva spôsobených elektrickým prúdom

  15. Accumulation of metallic elements into the superficial peat layer of mires and wet mineral soils of Estonian forest land / Akumulacja metali w powierzchniowej warstwie gleb torfowych i wilgotnych glebach mineralnych ekosystemów leśnych Estonii

    Directory of Open Access Journals (Sweden)

    Kõlli Raimo

    2015-12-01

    Full Text Available Z wykorzystaniem wody królewskiej badano stopień akumulacji metali (Al, Mg, Pb, Zn, Hg, Cd w epipedonie torfów wysokich (EP, torfów przejściowych, oglejonych glebach torfowych oraz zabagnionych glebach bielicowych. Współczynnik akumulacji metali (Kac w EP oszacowano w zależności od runa leśnego (FF, jako wkład i w porównaniu do głębszych poziomów (SS, jako tło przeszłości. Stwierdzono, że stopień akumulacji metali jest uzależniony of typu gleby (torfu oraz jej właściwości. W powstałym EP z FF stężenia Al i Pb zwiększyły się średnio 2.5-5.0 razy, natomiast Hg istotnie jedynie w glebach torfowych (średnio 1.5-1.6 razy. Istotnie zmniejszyły się (Kac 0.1-0.5 razy zawartości Zn we wszystkich glebach. Porównując zawartości metali w EP z SS stwierdzono znaczącą akumulację Pb, Cd, Zn i Hg w (~20 cm powierzchniowej warstwie torfowej.

  16. Enzymatyczna terapia zastępcza w leczeniu mukopolisacharydoz

    Directory of Open Access Journals (Sweden)

    Katarzyna Prusek

    2011-04-01

    Full Text Available Mukopolisacharydozy (MPS są grupą chorób będących wynikiemnadmiernego gromadzenia glikozoaminoglikanów w komórce(tab. I. Te makrocząsteczki w prawidłowych warunkach są rozkładanew lizosomach przez enzymy hydrolityczne. W przypadku brakulub upośledzonej aktywności hydrolaz lizosomalnych dochodzido nagromadzenia glikozoaminoglikanów, zwiększenia rozmiarówkomórki z następczym uszkodzeniem lub upośledzeniem funkcjonowaniaorganizmu. Wprowadzenie enzymatycznej terapii zastępczej(enzyme replacement therapy – ERT umożliwiło względnieskuteczne leczenie pacjentów z mukopolisacharydozami, którzydotychczas mogli być leczeni jedynie w sposób objawowy (tab. II.Obecnie ERT dostępna jest w przypadku MPS typu I, II i VI. W każdejz tych chorób okresowo drogą dożylną uzupełniany jest brakującyenzym. Udowodniono, iż ERT istotnie wpływa na poprawęjakości życia chorych, zmniejszając nasilenie większości objawów,podawany enzym nie przenika natomiast bariery krew–mózg, niewpływa zatem na zmiany neurologiczne.

  17. Současný stav a vývojové tendence v konstrukci hybridních pohonů pro osobní automobily

    OpenAIRE

    Štěnička, Petr

    2008-01-01

    Tato bakalářská práce pojednává o současném stavu a vývojových tendencích hybridních pohonů osobních automobilů. Obsahuje stručný historický přehled hybridních automobilů, jejich rozdělení, hlavní součásti a přehled současně vyráběných typů. Celá práce je zaměřena na popis principů tohoto typu pohonu a je zde také uvedeno srovnání výhod a nevýhod hlavních konstrukčních řešení. This bachelor thesis treats of the present state and development trends of hybrid propulsion of passenger cars. It...

  18. Ocelová konstrukce zastřešení hangáru

    OpenAIRE

    Pojezný, Tomáš

    2012-01-01

    Obsahem této práce je návrh a statické posouzení ocelové konstrukce zastřešení hangáru pro vrtulová letadla typu Cessna. Půdorysné rozměry jsou 40,0 x 60,0 m, výška objektu cca 12,0 m. Nosný systém konstrukce zastřešení je tvořen rovinnými příčnými vazbami z příhradových vazníků eliptického tvaru. Návrh konstrukce je volen ze dvou předběžně řešených geometrických a konstrukčních variant. Ty jsou porovnány na základě ceny, velikosti nátěrových ploch a hmotnosti konstrukce. Pro vybranou variant...

  19. Zróżnicowanie zawartości Cd, Pb, Zn i Cu w liściach tytoniu szlachetnego(Nicotiana tabacum l. uprawianegow rejonie Proszowic

    Directory of Open Access Journals (Sweden)

    Artur Szwalec

    2016-12-01

    Full Text Available Celem pracy była ocena zróżnicowania zawartość Cd, Pb, Zn i Cu w tytoniu typu Virginia uprawianym w rejonie Proszowic. Próby roślin i gleb pobierano z pól uprawnych położonych w sołectwach Bobin, Wolwanowice, Kościelec, Mysławczyce oraz Kuchary. Zawartości kadmu, ołowiu, cynku i miedzi oznaczono metodą FAAS na aparacie Solaar M6 firmy Unicam. Badane liście tytoniu cechowały się słabym stopniem kumulacji ołowiu, cynku i miedzi oraz intensywnym stopniem kumulacji kadmu. Stwierdzono dodatnie korelacje pomiędzy zawartościami kadmu i miedzi w glebie a stężeniami tych pierwiastków w liściach tytoniu. Zależność ta jest odwrotna (korelacja ujemna w przypadku cynku. Nie stwierdzono statystycznie istotnej korelacji w odniesieniu do zawartości ołowiu. Analiza wariancji wykazała zróżnicowanie zawartości badanych metali zarówno w liściach, jak i glebach.

  20. Robin Heart Surgery Robotic System. Challenges in Mechanical Construction, Control System and Stuff Training Before First Clinical Application

    Directory of Open Access Journals (Sweden)

    Nawrat Zbigniew

    2014-03-01

    Full Text Available Pojawienie sie w ostatnich dekadach na salach operacyjnych klinik zautomatyzowanych telemanipulatorów wprowadziło nowy standard w chirurgii małoinwazyjnej dzieki poprawie precyzji działania, powtarzalnosci ruchów i procedur, zwiekszenia komfortu pracy chirurga-zdalnego operatora - bardzo istotnego przy czesto długotrwałych zabiegach. Aby osiagnac ten etap wdrozenia klinicznego, biorac pod uwage z jednej strony wciaz istniejace ograniczenia tego typu zabiegów oraz koniecznosc spełnienia bardzo rygorystycznych wymagan certyfikacji samego produktu medycznego o najwyzszym w tym przypadku stopniu inwazyjnosci konieczne jest nie tylko opracowanie samego produktu finalnego spełniajacego wymagania norm lecz równiez certyfikacja całego procesu technologicznego jego wytwarzania. Celem pracy jest przedstawienie przygotowan do badan klinicznych, po fazie testów laboratoryjnych i na zwierzetach samego systemu robota, stanowisk testowych oraz przygotowania kadry dla jego obsługi dla projektu polskiego telemanipulatora chirurgicznego RobinHeart. Przedstawiony został projekt mechaniczny, systemu sterowania oraz stanowisk trenazerów-symulatorów i wybrane elementy modyfikacji podsystemów, powstałe na bazie 12 letnich doswiadczen zespołu i analizy rozwiazan swiatowych, dla najblizszego wdrozeniu robota RobinHeart Vision, przeznaczonego do zdalnej manipulacji torem wizyjnym podczas operacji małoinwazyjnych.

  1. Analytical And Numerical Solutions Of Metal High-Pressure Wave-Ring Gasket And Comparison With Experimental Results

    Directory of Open Access Journals (Sweden)

    Szybinski Bogdan

    2015-03-01

    Full Text Available W pracy przedstawiono wyniki badań doświadczalnych zestawu metalowych wysokociśnienio­wych uszczelek typu „B”, różniących się grubością i wciskiem montażowym. Badania wykonano w warunkach montażowych po wprowadzeniu uszczelek w gniazda, bez obciążenia połączenia ciśnieniem roboczym. Na wewnętrznej, cylindrycznej powierzchni uszczelek mierzono odkształce­nia obwodowe i osiowe przy użyciu tensometrów oporowych. Po odciążeniu i demontażu zmie­rzono szerokość strefy kontaktu plastycznie odkształconej powierzchni roboczej uszczelki. Zostały również wykonane badania materiałowe uszczelek oraz gniazd. Przeprowadzono numeryczną wery­fikację wyników pomiarów za pomocą obliczeń MES, w których uwzględniono nieliniowe własnoś­ci materiałów złącza, efekty kontaktowe oraz tarcie na powierzchni kontaktu. Dodatkowo porów­nano wyniki pomiarów z obliczeniami analitycznymi, otrzymanymi w oparciu o uproszczony powłokowy model uszczelki.

  2. Řízení 3D tiskárny pomocí mikropočítače Raspberry Pi

    OpenAIRE

    Lokajíček, Lukáš

    2015-01-01

    Předkládaná bakalářská práce se zabývá možností řešení ovládání 3D tiskárny typu RepRap pomocí mikropočítače Raspberry Pi. V teoretické části je provedena rešerše teorie o 3D tisku orientovaná na 3D tiskárnu na ústavu Elektrotechnologie VUT v Brně. Další část je pak věnována praktickému návrhu kompaktního rozhraní, které se bude starat o komunikaci mezi 3D tiskárnou a řídícím softwarem pomocí Raspberry Pi. The bachelor project deals with the possibility of solution to control 3D printer ty...

  3. Forced and gravity reflood experiments in hexagonal bundle lattices. Final report; Zwangs- und Schwerkraftflutversuche an hexagonalen Stabbuendelgeometrien; Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Wiehr, K.; Erbacher, F.; Just, W.; Harten, U.; Schmidt, H.

    1993-11-01

    This report describes reflooding experiments in two different fuel rod simulator bundles (FRSB) with hexagonal lattices in different bundle geometries and different pitch over diameter (p/d) ratios. The bundles consist of 16 electrically heated fuel rod simulators with a cosine shaped axial power profile. The tight bundle (p/d=1,06) has steel claddings with integrated helical fins as spacers, whereas the wider bundle (p/d=1,24) is equipped with grid spacers and with Zircaloy claddings typical of a convoy PWR. This report delivers mainly the data base for modelers in order to verify and further develop reflooding codes (e.g. ATHLET, RELAP) for hexagonal fuel lattices. Presently this is of immediate interest for the hexagonal fuel elements of Russian VVER`s and other advanced reactor concepts. The experimental data of a total of 45 experiments (most of them with forced feed and some of them with gravity feed) are stored on data tapes and data discs of the KfK IBM 3090. To find the proper experiments described for calculations, this report contains schematic figures of the test loop, the instrumentation of the bundle and the whole test section. The test parameters are listed in tables. To facilitate the identification of the experiments, the cladding temperatures of the inner 37 fuel rod simulators, the axial cladding temperatures and the total pressure drop of the bundle are plotted. A user description for the experimental data and the corresponding measuring channel list of each experiment enables to find the proper test. In the appendix the material properties to model the rod simulator are given. (orig.) [Deutsch] Dieser Bericht beschreibt Flutexperimente an zwei Brennstabsimulatorbuendeln mit hexagonaler Stabanordnung unterschiedlicher Buendelgeometrie. Die Buendel bestehen aus jeweils 61 Brennstabsimulatoren (BSS) mit einem cosinusfoermigen axialen Leistungsprofil, wobei das sehr enge Buendel (p/d=1,06) mit Stahlhuellen mit integrierten Wendelrippen als

  4. Progress and perspectives of ASTEC applications in the European Network SARNET

    Energy Technology Data Exchange (ETDEWEB)

    Van Dorsselaere, J.P. [Institut de Radioprotection et de Surete Nucleaire (IRSN/DPAM), 13 - St-Paul-Lez-Durance (France); Allelein, H.J.; Neu, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koln (Germany)

    2006-07-01

    The ASTEC integral code is jointly developed by IRSN (France) and GRS (Germany) for LWR source term Severe Accident (SA) evaluation, PSA level 2 studies and SA management evaluation. ASTEC constitutes now the reference European integral code through its role in the Network SARNET (Severe Accident Network of Excellence) in the EC 6. Framework Program. The models of next version V1.3, released before end of 2006, represent the current State of the Art, its validation is very extensive (in particular on Phebus FP) and, after next implementation of a model for reflooding of degraded cores, it will cover all needs for SA evaluation in PWR and VVER. It will be the reference code for the IRSN PSA level 2 (Probabilistic Safety Analysis) on French PWR 1300 MWe that starts in 2006. In the frame of SARNET, IRSN coordinates the ASTEC Topic gathering 30 partners that assess the code through validation against experiments and benchmarks with reference codes like CATHARE or RELAP5 for the reactor coolant circuit and COCOSYS for the containment. Plant application calculations are compared with MELCOR and MAAP4 results for a series of different SA sequences. Besides, the knowledge generated by SARNET Topics (Corium, Source Term and Containment) will be progressively integrated into the code through improved or new models. The 2. Users' Club organized at Aix-en-Provence in June 06, with 45 participants from 27 organizations, allowed fruitful discussions with the Maintenance Team. After 2 years of work, code validation shows good overall results, often close to results of reference codes. Some results reach the limits of present knowledge, for instance on Molten-Corium-Concrete-Interaction (MCCI) and Direct Containment Heating (DCH). Benchmarks on plant applications have been performed on diverse reactor types: PWR 900, Konvoi 1300, Westinghouse 1000, VVER-1000 and VVER-440. The main trends of results are similar to MELCOR or MAAP4 results. The objective of the quantitative

  5. Uncertainty and sensitivity analysis for the modeling of transients with interaction of thermal hydraulics and neutron kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Soeren Kliem; Siegfried Mittag [Forschungszentrum Rossendorf (FZR), Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Siegfried Langenbuch [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH, P.O.B. 13 28, D-85748 Garching (Germany)

    2005-07-01

    Full text of publication follows: The transition from the application of conservative models to the use of best-estimate models raises the question about the uncertainty of the obtained results. This question becomes especially important, if the best-estimate models should be used for safety analyses in the field of nuclear engineering. Different methodologies were developed to assess the uncertainty of the calculation results of computer simulation codes. One of them is the methodology developed by Gesellschaft fuer Anlagenund Reaktorsicherheit (GRS) which uses the statistical code package SUSA. In the past, this methodology was applied to the calculation results of the advanced thermal hydraulic system code ATHLET. In the frame of the recently finished EU FP5 funded research project VALCO, that methodology was extended and successfully applied to different coupled code systems, including the uncertainty analysis for neutronics. These code systems consist of a thermal hydraulic system code and a 3D neutron kinetic core model. One of the code systems applied was ATHLET coupled with the Rossendorf kinetics code DYN3D. Two real transients at NPPs with VVER-type reactors documented within the VALCO project were selected for analyses. One was the load drop of one of two turbines to house load level at the Loviisa-1 NPP (VVER-440), the second was a test with the switching-off of one of two main feed water pumps at the VVER-1000 Balakovo-4 NPP. The current paper is dedicated to the different steps of the use and implementation of the GRS methodology to coupled code systems and to the assessment of the results obtained by the DYN3D/ATHLET code. Based on the relevant physical processes in both transients, lists of possible sources of uncertainties were compiled. They are specific for the two transients. Besides control parameters like control rod movement and thermal hydraulic parameters like secondary side pressure, mass flow rates, pressurizer sprayer and heater

  6. Zaburzenia lipidowe u pacjentów z zawrotami głowy

    Directory of Open Access Journals (Sweden)

    Hanna Zielińska-Bliźniewska

    2012-11-01

    Full Text Available Wprowadzenie: Celem pracy była ocena zaburzeń lipidowych u pacjentów z zawrotami głowy. Materiał i metody: Badania przeprowadzono na grupie 918 chorych, w tym 598 kobiet i 320 mężczyzn, w wieku 18–83 lat (średnia wieku 55±0,5, leczonych w latach 2009–2011 w Klinice Otolaryngologii i Onkologii Laryngologicznej z Zespołem Pracowni Audiologicznych i Foniatrycznych Uniwersyteckiego Szpitala Klinicznego im. WAM w Łodzi z powodu zawrotów głowy. U wszystkich chorych przeprowadzono szczegółowy wywiad, badanie przedmiotowe otolaryngologiczne, otoneurologiczne. Każdy pacjent był konsultowany neurologicznie, okulistycznie i internistycznie oraz miał wykonywane USG naczyń doczaszkowych, tomografię komputerową odcinka szyjnego kręgosłupa i głowy w celu wykluczenia schorzeń organicznych ośrodkowego układu nerwowego. Przeprowadzono także badania laboratoryjne, takie jak stężenie cholesterolu całkowitego, triglicerydy, frakcję cholesterolu LDL i HDL oraz stężenie glukozy w surowicy krwi. Wyniki: W grupie 918 pacjentów z zawrotami głowy u 539 (58,71% miały one pochodzenie ośrodkowe, a u 379 chorych (41,28% charakter mieszany, w tym u 366 kobiet (67,90% rozpoznano zawroty pochodzenia ośrodkowego, a u 232 (61,21% typu mieszanego. Spośród 320 mężczyzn (34,78% z zawrotami głowy u 173 (32,09% stwierdzono zawroty pochodzenia ośrodkowego, a u 147 (38,78% typu mieszanego. Analizując stężenia frakcji lipidów u badanych, odnotowano podwyższone wartości cholesterolu całkowitego u 67,03% z nich, w tym u 71,34% mężczyzn i 64,76% kobiet. Podwyższone stężenia frakcji cholesterolu LDL zaobserwowano u 51,57% pacjentów, w tym u 54,83% mężczyzn i 49,83% kobiet. Frakcja HDL cholesterolu u większości chorych (61,99% była w normie. Również stężenie triglicerydów u większości badanych (u 69,45% nie odbiegało od normy, podobnie jak stężenie glukozy (u 59,25% mężczyzn oraz 67,78% kobiet. Wnioski: Zaburzenia

  7. Znaczenie witaminy D w chorobach atopowych u dzieci

    Directory of Open Access Journals (Sweden)

    Agnieszka Rustecka

    2013-04-01

    Full Text Available Alergia stanowi istotny problem w dzisiejszym świecie. Często jej podłożem jest atopia, czyli uwarunkowana genetycznie predyspozycja do nadmiernej produkcji przeciwciał klasy IgE. Do najczęstszych chorób alergicz‑ nych wieku rozwojowego należą: atopowe zapalenie skóry, astma, alergiczny nieżyt nosa, alergiczny nieżyt spo‑ jówek, alergie pokarmowe i pokrzywki. Udowodniony wpływ na rozwój chorób atopowych mają czynniki gene‑ tyczne, zakażenia wirusowe oraz czynniki środowiskowe. Ostatnio duże zainteresowanie wzbudza rola niedoboru witaminy D (cholekalcyferolu w rozwoju tego rodzaju schorzeń. Witamina D oraz jej aktywne meta‑ bolity i syntetyczne analogi są tradycyjnie stosowane w postaci preparatów doustnych w celu regulowania homeostazy wapniowo-fosforanowej. Witamina D, wytwarzana w skórze z prekursora steroidowego, jest nie‑ aktywnym biologicznie prohormonem. W organizmie ulega dwustopniowej enzymatycznej hydroksylacji (w wątrobie i nerkach do utworzenia głównej aktywnej formy hormonalnej. Biologicznie najwyższą aktywność wykazuje metabolit 1,25(OH2D. W celach diagnostycznych oznacza się stężenie 25(OHD i 1,25(OH2D. Badania wykonane na dużych populacjach wskazują na ścisłą korelację pomiędzy długotrwałym niedoborem witaminy D a zaburzeniami metabolicznymi. Niedobór cholekalcyferolu w surowicy zwiększa ryzyko rozwoju choroby o podłożu autoimmunologicznym lub infekcyjnym. Jego działanie, wykraczające poza gospodarkę mineralną, wynika z pobudzenia pozanerkowego receptora dla witaminy D (VDR. Sugeruje się związek między spożyciem witaminy D przez kobiety w ciąży a ryzykiem wystąpienia chorób atopowych u ich dzieci. W świetle ostatnich badań cholekalcyferol jest czynnikiem immunoprotekcyjnym w chorobach autoimmunologicznych, ogranicza odpowiedź immunologiczną typu Th1 i przesuwa ją w stronę dominacji odpowiedzi typu Th2. Z niektórych przytoczonych publikacji wynika, że w

  8. Confined Phase Envelope of Gas-Condensate Systems in Shale Rocks

    Science.gov (United States)

    Nagy, Stanislaw; Siemek, Jakub

    2014-12-01

    Natural gas from shales (NGS) and from tight rocks are one of the most important fossil energy resource in this and next decade. Significant increase in gas consumption, in all world regions, will be marked in the energy sector. The exploration of unconventional natural gas & oil reservoirs has been discussed recently in many conferences. This paper describes the complex phenomena related to the impact of adsorption and capillary condensation of gas-condensate systems in nanopores. New two phase saturation model and new algorithm for search capillary condensation area is discussed. The algorithm is based on the Modified Tangent Plane Criterion for Capillary Condensation (MTPCCC) is presented. The examples of shift of phase envelopes are presented for selected composition of gas-condensate systems. Gaz ziemny z łupków (NGS) oraz z ze złóż niskoprzepuszczalnych (typu `tight') staje się jednym z najważniejszych zasobów paliw kopalnych, w tym i następnym dziesięcioleciu. Znaczący wzrost zużycia gazu we wszystkich regionach świata zaznacza się głównie w sektorze energetycznym. Rozpoznawanie niekonwencjonalnych złóż gazu ziemnego i ropy naftowej w ostatnim czasie jest omawiane w wielu konferencjach. Niniejszy artykuł opisuje złożone zjawiska związane z wpływem adsorpcji i kapilarnej kondensacji w nanoporach w złożach gazowo-kondensatowych. Pokazano nowy dwufazowy model równowagowy dwufazowy i nowy algorytm wyznaczania krzywej nasycenia w obszarze kondensacji kapilarnej. Algorytm bazuje na kryterium zmodyfikowanym płaszczyzny stycznej dla kapilarnej kondensacji (MTPCCC). Przykłady zmiany krzywych nasycenia są przedstawiane w wybranym składzie systemów gazowo- kondensatowych

  9. Pozatwarzowy ziarniniak twarzy – opis przypadku

    Directory of Open Access Journals (Sweden)

    Lilianna Kulczycka

    2010-09-01

    Full Text Available Wprowadzenie: Ziarniniak twarzy (granuloma faciale jest nieczęstospotykaną, przewlekłą chorobą skóry. Charakteryzuje się występowaniemna twarzy brunatnoczerwonych grudek, guzków i ognisk naciekowych.Wyjątkowo ziarniniaka spotyka się poza twarzą. Cel pracy: Przedstawienie chorego, u którego granuloma faciale był zlokalizowanywyłącznie poza twarzą. Opis przypadku: U 56-letniego, ogólnie zdrowego mężczyzny w ostatnich15 latach stopniowo ukazywały się bezobjawowe, pojedynczegrudki i ograniczone, brunatnoczerwone, koliste nacieki, o średnicy1–4 cm. Wykwity były zlokalizowane na plecach, barku, brzuchu i zalewą małżowiną uszną. Wymagały różnicowania z chłoniakiem, lymphocytoma,mięsakiem Kaposiego oraz sarkoidozą. W badaniu histopatologicznymstwierdzono obraz wskazujący na ziarniniaka twarzy, copozwoliło na ustalenie rozpoznania granuloma faciale o lokalizacji pozatwarzowej.Po 6-miesięcznej terapii dapsonem i takrolimusem stosowanymmiejscowo wykwity ustąpiły, pozostawiając jedynie przebarwienia. Wnioski: W różnicowaniu przewlekłych zmian skórnych typu brunatnoczerwonychgrudek, guzków oraz wyniosłych nacieczonych ognisk,niezależnie od ich umiejscowienia, należy brać pod uwagę rozpoznaniepozatwarzowego ziarniniaka twarzy. W diagnostyce decydująceznaczenie ma badanie histopatologiczne. Korzystnym leczeniem jestogólne stosowanie dapsonu oraz miejscowo takrolimusu.

  10. Twardzina ograniczona pęcherzowa – opis przypadku

    Directory of Open Access Journals (Sweden)

    Elżbieta Meszyńska

    2011-01-01

    Full Text Available Wprowadzenie: Pęcherze pojawiające się w obrębie zmian typu twardzinyograniczonej są rzadko opisywane. Etiologia ich powstawaniajest wieloczynnikowa. Najczęściej rozważany jest ucisk blaszek twardzinowychna naczynia limfatyczne. Twardzina pęcherzowa wymagaróżnicowania z liszajem twardzinowym i zanikowym, w którym u częścichorych obserwuje się występowanie pęcherzy. Cel pracy: Przedstawienie przypadku 41-letniej kobiety z twardzinąograniczoną (morphea, u której w 15. roku trwania choroby wystąpiłypęcherze w obrębie ognisk twardzinowych. Opis przypadku: U 41-letniej kobiety z trwającymi od 15 lat ogniskamimorphea wystąpiły nadżerki oraz pęcherze krwotoczne w obrębieognisk stwardniałych. Badanie histopatologiczne wykazało obrzękskóry właściwej oraz obecność pęcherzy podnaskórkowych. Badaniaimmunologiczne (metoda immunofluorescencji pośredniej i bezpośredniejbyły negatywne. Obraz kliniczny i histopatologiczny pozwoliłna rozpoznanie odmiany pęcherzowej twardziny ograniczonej (morpheabullosa. Zmiany pęcherzowe ustąpiły po 40 dniach leczeniacefalosporynami (cefuroksym. Wnioski: Morphea bullosa jest rzadką odmianą twardziny skórnej, którawymaga różnicowania z liszajem twardzinowym i zanikowym orazautoimmunologicznymi, podnaskórkowymi chorobami pęcherzowymi.Wykwity pęcherzowe w obrębie ognisk morphea mogą wystąpićnawet po wielu latach utrzymywania się typowych blaszek stwardnieniowych.

  11. Analysis of Network Traffic Filtering / Analiza Filtracji Ruchu Sieciowego

    Directory of Open Access Journals (Sweden)

    Kabala Piotr

    2015-09-01

    Full Text Available Artykuł przedstawia badania koncepcji Platformy Sieciowej z zaporą Linux PLD wykorzystującą dostępne narzędzia analizy ruchu, implementację wybranego mechanizmu bezpieczeństwa oraz analizę efektywności przygotowanej Zapory Sieciowej (FW. Badania oparto na architekturze ekranowanych podsieci. Testowanie skuteczności zapory wykonano w dwóch etapach. W pierwszej części sprawdzono bezpieczeństwo poprzez testowanie odporności na próby skanowania portów i najczęstsze typy ataku. Druga część badań miała na celu pokazanie wpływu zapory na czas i opóźnienie realizacji transakcji dla ruchu HTTP w zależności od poziomu bezpieczeństwa i natężenia ruchu generowanego przez użytkowników sieci wewnętrznej. Sprawdzenie skuteczność FW zostało zweryfikowane przez badania odporności w próbie skanowania portów i ataków odmowy usługi (tj. SYN Flood i ICMP zapewniając bazę rozproszonych ataków typu DDoS. Ta część badań miała jeden cel, którym było pokazanie wpływu FW na czas i opóźnienia realizacji transakcji dla ruchu HTTP, w zależności od poziomu bezpieczeństwa i intensywności ruchu generowanego przez użytkowników sieci.

  12. A Visual mining based framework for classification accuracy estimation

    Science.gov (United States)

    Arun, Pattathal Vijayakumar

    2013-12-01

    Classification techniques have been widely used in different remote sensing applications and correct classification of mixed pixels is a tedious task. Traditional approaches adopt various statistical parameters, however does not facilitate effective visualisation. Data mining tools are proving very helpful in the classification process. We propose a visual mining based frame work for accuracy assessment of classification techniques using open source tools such as WEKA and PREFUSE. These tools in integration can provide an efficient approach for getting information about improvements in the classification accuracy and helps in refining training data set. We have illustrated framework for investigating the effects of various resampling methods on classification accuracy and found that bilinear (BL) is best suited for preserving radiometric characteristics. We have also investigated the optimal number of folds required for effective analysis of LISS-IV images. Techniki klasyfikacji są szeroko wykorzystywane w różnych aplikacjach teledetekcyjnych, w których poprawna klasyfikacja pikseli stanowi poważne wyzwanie. Podejście tradycyjne wykorzystujące różnego rodzaju parametry statystyczne nie zapewnia efektywnej wizualizacji. Wielce obiecujące wydaje się zastosowanie do klasyfikacji narzędzi do eksploracji danych. W artykule zaproponowano podejście bazujące na wizualnej analizie eksploracyjnej, wykorzystujące takie narzędzia typu open source jak WEKA i PREFUSE. Wymienione narzędzia ułatwiają korektę pół treningowych i efektywnie wspomagają poprawę dokładności klasyfikacji. Działanie metody sprawdzono wykorzystując wpływ różnych metod resampling na zachowanie dokładności radiometrycznej i uzyskując najlepsze wyniki dla metody bilinearnej (BL).

  13. Nabyta hemofilia A u pacjentki z reumatoidalnym zapaleniem stawów

    Directory of Open Access Journals (Sweden)

    Stanisława Bazan-Socha

    2011-12-01

    Full Text Available Nabyta hemofilia jest rzadkim, potencjalnie śmiertelnym, nabytymschorzeniem autoimmunologicznym, z wytworzeniem przeciwciałblokujących aktywność własnych czynników krzepnięcia. Toobniżenie aktywności prowadzi do rozwoju skazy krwotocznej.Najczęściej wytwarzane są przeciwciała przeciwko czynnikowi VIII– określane jako nabyta hemofilia A. W blisko połowie przypadkówma ona charakter pierwotny i nie towarzyszy żadnym innym zmia -nom patologicznym, w 35–40% jest wtórna do innych schorzeńautoimmunologicznych, nowotworowych lub reakcji polekowych,a w 10–15% dotyczy młodych kobiet w okresie połogu. Leczenienabytej hemofilii polega na opanowaniu czynnego krwawienia orazeliminacji z krążenia aktywnego inhibitora.Autorzy przedstawiają przypadek 55-letniej chorej na reumatoidalnezapalenie stawów z nabytą hemofilią typu A (ryc. 1 i 2. Objawy skazykrwotocznej ustąpiły po dożylnym podaniu rekombinowanegoaktyw nego czynnika VII oraz prednizonu i cyklofosfamidu. Celem pracyjest zwrócenie uwagi na występowanie tego schorzenia u pacjentówz chorobami reumatycznymi. Nagłe wystąpienie objawów skazykrwotocznej, a szczególnie krwawień podskórnych, połączonez wydłużeniem czasu częściowej tromboplastyny po aktywacjipowinno budzić podejrzenie nabytej hemofilii. Życie choremu możeuratować jedynie szybka diagnostyka i właściwe leczenie.

  14. MATERIA ORGANICZNA W WODACH NATURALNYCH – FORMY WYSTĘPOWANIA I METODY OZNACZANIA

    Directory of Open Access Journals (Sweden)

    Andżelika PIETRZYK

    2016-06-01

    Full Text Available Ilość materii organicznej jest ważnym parametrem decydującym o stopniu zanieczyszczenia wód naturalnych. Jedynym dobrze zdefiniowanym wskaźnikiem określającym sumę wszystkich organicznych zanieczyszczeń jest ogólny węgiel organiczny (OWO. W artykule omówione zostały metody wykorzystywane do pomiaru OWO oraz formy występowania materii organicznej w różnych typach wód. Zawartość ogólnego węgla organicznego w wodach naturalnych jest zróżnicowana i zależy od następujących czynników: typu i wielkości badanego zbiornika wodnego, położenia geograficznego, temperatury, zasolenia, wartości pH, aktywności mikrobiologicznej oraz charakteru zlewni. Wzrost zawartości ogólnego węgla organicznego w wodach, które ujmowane są w celu zaopatrywania ludności w wodę przeznaczoną do picia wpływa na znaczne zwiększenie kosztów jej uzdatniania. Obecność substancji organicznych, a przede wszystkim substancji humusowych przyczynia się do pogorszenia właściwości organoleptycznych wody m.in. są one odpowiedzialne za występowanie specyficznego smaku i zapachu, a także za ponadnormatywną barwę.

  15. Styles of Humor of Charges and Educators and the Social Climate of Juvenile Correctional Institutions

    Directory of Open Access Journals (Sweden)

    Anna Karłyk- Ćwik

    2017-01-01

    Full Text Available Autorka prezentuje wyniki badań empirycznych, których celem było zdiagnozowanie klimatu społecznego instytucji resocjalizacyjnych dla nieletnich oraz opisanie stylów humoru prezentowanych przez wychowawców i wychowanków tych placówek, a także ustalenie wzajemnych relacji pomiędzy tymi zmiennymi. Badaniami kwestionariuszowymi (z wykorzystaniem Skali Klimatu Społecznego i Kwestionariusza Stylów Humoru objęto grupę 162 wychowanków oraz 52 wychowawców z czterech Młodzieżowych Ośrodków Wychowawczych. Przeprowadzone badania ujawniły, że klimat społeczny tych placówek jest najbardziej zbliżony do typu „opiekuńczo-wychowawczego”, a w funkcjonowaniu intrapsychicznym i interpersonalnym, zarówno wychowawców, jak i wychowanków, adaptacyjne style humoru przeważają nad stylami nieprzystosowawczymi. Ponadto ustalono, że to wychowankowie, za sprawą prezentowanych stylów humoru, wydają się mieć większy wpływ na kreowanie klimatu społecznego placówek, w których przebywają, niż personel resocjalizacyjny. Analiza uzyskanych wyników w kontekście koncepcji resilience zachęca do traktowania humoru w kategoriach „płaszczyzny oporu” umożliwiającej młodzieży niedostosowanej społecznie „odbicie się od dna”.

  16. ZASTOSOWANIE FUNKCJONAŁU HU-WASHIZU W PLASTYCZNEJ ANALIZIE MES PŁYT GRUBYCH

    Directory of Open Access Journals (Sweden)

    Jakub LEWANDOWSKI

    2016-07-01

    Full Text Available W pracy sformułowano oryginalny, autorski funkcjonał dla zagadnień teorii plastyczności. Podstawą był funkcjonał Hu-Washizu z teorii sprężystości. Przyrostowa postać funkcjonału pozwala w prosty sposób budować algorytmy MES. Zastosowanie funkcjonału przedstawiono na przykładzie płyty grubej. Zastosowano model warstwowy aby uwzględnić częściowe uplastycznienie przekroju płyty. Algorytm MES dla płyty grubej zbudowano w oparciu o trójkątny trzy węzłowy element skończony z liniowymi funkcjami kształtu dla wszystkich przemieszczeń uogólnionych. Naprężenia i odkształcenia w tego typu elemencie przyjmuje się jako stałe. Przedstawiony algorytm nie wymaga żadnych dodatkowych równań teorii plastyczności i jest równoważny stowarzyszonemu prawu płynięcia plastycznego. Algorytm prowadzi do nieliniowego, przyrostowego układu równań algebraicznych, który rozwiązuje się metodą Newtona. Kilka prostych przykładów pozytywnie weryfikuje przyjęte założenia i stosowane algorytmy.

  17. Hydroxycinnamic derivatives content in plant organs linked to harvest time of Salvia officinalis L. cv. ‘Krajová’ / Obsah hydroxyškoricových derivátov v rastlinných orgánoch Salvia officinalis L. cv. ‘Krajová’ v závislosti od termínu zberu

    Directory of Open Access Journals (Sweden)

    Tekeľová D.

    2015-06-01

    Full Text Available Salvia officinalis L. (šalvia lekarska je vyznamnou silicovou liečivou rastlinou domacou v oblasti Stredomoria, pre farmaceuticke učely sa pestuje. Okrem silice sa na biologickom učinku rastliny podieľaju hlavne diterpeny, triterpeny a fenolove latky typu hydroxyškoricovych derivatov a flavonoidov. Je zname, že obsah a kvalita silice v šalvii koliše v zavislosti od rastlinnej časti, vyvinovej fazy, klimatickych a podnych podmienok. V našej praci sme sledovali kolisanie obsahu celkovych hydroxyškoricovych derivatov (THD a samotnej kyseliny rozmarinovej (RA v nadzemnych častiach šalvie lekarskej v roznych terminoch zberu. Obsah THD v suchych listoch (Salviae officinalis folium stanoveny liekopisnou metodou kolisal v jednotlivych terminoch zberu od 3,06 % do 3,52 %, najvyšši bol v listoch z vyhonkov najmladšich rastlin a z novonarastenych vyhonkov v septembri. Podobne kolisanie obsahu THD v jednotlivych zberoch bolo aj v stonkach, tie však obsahovali len 1,33 - 3,04 %. Rovnaku variabilitu obsahu sme zaznamenali pri kyseline rozmarinovej, jej obsah v listoch kolisal od 0,76 % do 1,65 % a v stonkach od 0,19 % do 0,83 %. Najvyšši obsah THD a RA bol vo vrcholovych listoch, najnižši v listoch umiestnenych v strede stonky. Počas kvitnutia rastliny sa obsah THD a RA v listoch znižil.

  18. Zmiany czynności kory ruchowej mózgu po leczeniu botuliną u pacjentów ze stwardnieniem rozsianym i spazmem kończyn dolnych

    Directory of Open Access Journals (Sweden)

    Pavel Hok

    2011-12-01

    Full Text Available Miejscowe skurcze toniczne to powszechnie spotykany objaw stwardnienia rozsianego (łac. sclerosis multiplex, SM. Do ich zwalczania coraz częściej stosowany jest zastrzyk domięśniowy botuliny typu A. Do analizy statystycznej zaakceptowaliśmy 4 z 10 badanych pacjentów z SM i spastycznością kończyn dolnych oraz 5 zdrowych wolontariuszy. Pacjenci zostali poddani badaniu fMRI trzykrotnie: w tygodniu przed zastrzykiem botuliny A, a następnie w 4. i 12. tygodniu po iniekcji. Podczas badań fMRI probanci wykonywali zginanie i prostowanie stawu kolanowego według planu blokowego, przy czym faza czynna zamieniała się z fazą spoczynku w 15-sekundowych odstępach. Obraz przeciętnej aktywacji pacjentów podczas pierwszej sesji wskazywał, w porównaniu z grupą kontrolną, na istotny wzrost aktywacji obustronnej kory czuciowo-ruchowej płatu czołowego i ciemieniowego. Podczas drugiej sesji w 4. tygodniu aktywacja zmalała do tego stopnia, że statystycznie nie różniła się od zdrowej kontroli. Z kolei w obrazach trzeciej sesji po 12 tygodniach odnotowano w związku z wygaśnięciem efektu botuliny A ponowny wzrost aktywacji niemal do objętości pierwotnej. Wnioski: Stwierdzamy, że aktywacja kory ruchowej odzwierciedla zmiany w obwodowym układzie nerwowym zachodzące podczas leczenia za pomocą botuliny A, w czym prawdopodobnie pośredniczą zmiany w aferentacji. Jest to nowe odkrycie, aczkolwiek nie wykracza poza stwierdzenia podobnych badań przeprowadzonych innymi metodami.

  19. Presentation of the results for deuterium retention and thermal release in a new type of low activation ferritic-martensitic steel EUROFER / Результаты исследования по удержанию дейтерия и термической десорбции в условиях низкой активации ферритно-мартенситной стали EUROFER / Rezultati zadržavanja i termalne desorpcije deuterijuma u EUROFER-u, novoj vrsti feritno-martenzitnog čelika niske aktivacije

    Directory of Open Access Journals (Sweden)

    Sanja Lj. Korica

    2016-04-01

    žđu, hromu i EOROFER-u, leguri koja se razmatra kao najnoviji materijal za buduće fuzione reaktore. Studija je pokazala sledeće rezultate: zadržavanje deuterijuma u hromu je mnogo veće nego u gvožđu (usled formiranja hidrida hroma, zadržavanje deuterijuma u EUROFER-u je za faktor 2 veće nego u gvožđu, primećena je specifična struktura u koncentracionom profilu gvožđa i EUROFER-a na dubini ~ 4 μm, veliki stepen difuznosti i zadržavanja deuterijuma govore o potencijalnoj upotrebi Au kao difuzione barijere u fuzionom reaktoru.

  20. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    Directory of Open Access Journals (Sweden)

    Weimin Ma

    2016-03-01

    Full Text Available A historical review of in-vessel melt retention (IVR is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs. The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

  1. Monte-Carlo Simulation of the Features of Bi-Reactior Accelerator Driven Systems

    CERN Document Server

    Bznuni, S A; Khudaverdian, A G; Barashenkov, V S; Sosnin, A N; Polyanskii, A A

    2002-01-01

    Parameters of accelerator-driven systems containing two "cascade" subcritical assemblies (liquid metal fast reactor, used as a neutron booster, and a thermal reactor, where main heat production is taking place) are investigated. Three main reactor cores analogous to VVER-1000, MSBR-1000 and CANDU-6 reactors are considered. Functioning in a safe mode (k_{eff}=0.94-0.98) these systems under consideration demonstrate much larger capacity in the wide range of k_{eff} in comparison with analogous systems without intermediate fast booster reactor and simultaneously having the density of thermal neutron flux equal to Phi^{max}=10^{14} cm^{-2}c^{-1} and operating with the fast and thermal zones they are capable to transmute the whole scope of nuclear waste reducing the requirements on the beam current of the accelerator by one order of magnitude. It seems to be the most important in case when molten salt thermal breeder reactor cores are considered as a main heat generating zone.

  2. Analysis of Steam Generators Corrosion Products from Slovak NPP Bohunice

    Directory of Open Access Journals (Sweden)

    Jarmila Degmová

    2012-01-01

    Full Text Available One of the main goals of the nuclear industry is to increase the nuclear safety and reliability of nuclear power plants (NPPs. As the steam generator (SG is the most corrosion sensitive component of NPPs, it is important to analyze the corrosion process and optimize its construction materials to avoid damages like corrosion cracking. For this purpose two different kinds of SGs and its feed water distributing systems from the NPP Jaslovske Bohunice were studied by nondestructive Mössbauer spectroscopy. The samples were scraped from the surface and analyzed in transmission geometry. Magnetite and hematite were found to be the main components in the corrosion layers of both SGs. Dependant of the material the SG consisted of, and the location in the system where the samples were taken, the ratios between magnetite and hematite and the paramagnetic components were different. The obtained results can be used to improve corrosion safety of the VVER-440 secondary circuit as well as to optimize its water chemistry regime.

  3. Verification of three-dimensional neutron kinetics model of TRAP-KS code regarding reactivity variations

    Energy Technology Data Exchange (ETDEWEB)

    Uvakin, Maxim A.; Alekhin, Grigory V.; Bykov, Mikhail A.; Zaitsev, Sergei I. [EDO ' GIDROPRESS' , Moscow Region, Podolsk (Russian Federation)

    2016-09-15

    This work deals with TRAP-KS code verification. TRAP-KS is used for coupled neutron and thermo-hydraulic process calculations of VVER reactors. The three-dimensional neutron kinetics model enables consideration of space effects, which are produced by energy field and feedback parameters variations. This feature has to be investigated especially for asymmetrical multiplying variations of core properties, power fluctuations and strong local perturbation insertion. The presented work consists of three test definitions. First, an asymmetrical control rod (CR) ejection during power operation is defined. This process leads to fast reactivity insertion with short-time power spike. As second task xenon oscillations are considered. Here, small negative reactivity insertion leads to power decreasing and induces space oscillations of xenon concentration. In the late phase, these oscillations are suppressed by external actions. As last test, an international code comparison for a hypothetical main steam line break (V1000CT-2, task 2) was performed. This scenario is interesting for asymmetrical positive reactivity insertion by decreasing coolant temperature in the affected loop.

  4. Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S.

    2001-01-11

    Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

  5. Gamma spectrometric characterization of short cooling time nuclear spent fuels using hemispheric CdZnTe detectors

    CERN Document Server

    Lebrun, A; Szabó, J L; Arenas-Carrasco, J; Arlt, R; Dubreuil, A; Esmailpur-Kazerouni, K

    2000-01-01

    After years of cooling, nuclear spent fuel gamma emissions are mainly due to caesium isotopes which are emitters at 605, 662 and 796-801 keV. Extensive work has been done on such fuels using various CdTe or CdZnTe probes. When fuels have to be measured after short cooling time (during NPP outage) the spectrum is much more complex due to the important contributions of niobium and zirconium in the 700 keV range. For the first time in a nuclear power plant, four spent fuels of the Kozloduy VVER reactor no 4 were measured during outage, 37 days after shutdown of the reactor. In such conditions, good resolution is of particular interest, so a 20 mm sup 3 hemispheric crystal was used with a resolution better than 7 keV at 662 keV. This paper presents the experimental device and analyzes the results which show that CdZnTe commercially available detectors enabled us to perform a semi-quantitative determination of the burn-up after a short cooling time. In addition, it is discussed how a burn-up evolution code (CESAR)...

  6. Nuclear safety in EU candidate countries

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

    Nuclear safety in the candidate countries to the European Union is a major issue that needs to be addressed in the framework of the enlargement process. Therefore WENRA members considered it was their duty to offer their technical assistance to their Governments and the European Union Institutions. They decided to express their collective opinion on nuclear safety in those candidate countries having at least one nuclear power plant: Bulgaria, the Czech Republic, Hungary, Lithuania, Romania, Slovakia and Slovenia. The report is structured as follows: A foreword including background information, structure of the report and the methodology used, General conclusions of WENRA members reflecting their collective opinion, For each candidate country, an executive summary, a chapter on the status of the regulatory regime and regulatory body, and a chapter on the nuclear power plant safety status. Two annexes are added to address the generic safety characteristics and safety issues for RBMK and VVER plants. The report does not cover radiation protection and decommissioning issues, while safety aspects of spent fuel and radioactive waste management are only covered as regards on-site provisions. In order to produce this report, WENRA used different means: For the chapters on the regulatory regimes and regulatory bodies, experts from WENRA did the work. For the chapters on nuclear power plant safety status, experts from WENRA and from French and German technical support organisations did the work. Taking into account the contents of these chapters, WENRA has formulated its general conclusions in this report.

  7. Short review: Potential impact of delamination cracks on fracture toughness of structural materials

    Directory of Open Access Journals (Sweden)

    X.C. Arnoult

    2016-02-01

    Full Text Available The current energy policy envisages extended lifetime for the current nuclear power plants (GEN II NPP. This policy imposes a large research effort to understand the ageing of power plant components. In this goal, it is necessary to improve knowledge about safety, reliability and components’ integrity for more than forty years of operation. In Central and Eastern Europe, the majority of NPPs are VVER types, where some of the components are produced from austenitic steel 08Ch18N10T. Irradiated 08Ch18N10T may exhibit brittle behavior, namely delamination cracks are found in some cases on the fracture surface of irradiated 08Ch18N10T with elongated δ-ferrite. Delamination cracks have also been observed on the fracture surface of high-strength steels or aluminum-lithium alloys. This article presents a state-of-the art review to provide a detailed analysis of the influence of delamination cracks on the toughness of metal alloys. In general, the delamination cracks are present in metal alloys having a high texture and microstructure anisotropy. Three types of delamination cracks have been observed and are classified as crack arrester delamination, crack divider delamination and crack splitting delamination. The microscopy characterization, 3D fracture theories and computational studies explaining possible causes and effects of delamination cracks on the mechanical properties of metal alloys are presented.

  8. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  9. A Comparison of Flow Field Characteristics from PIV Experiment Measurement to Numerical Simulation behind a Spacer in a Vertical Pipe

    Directory of Open Access Journals (Sweden)

    Lávička D.

    2010-07-01

    Full Text Available This paper describes the topic of measurement using a modern laser method (PIV in an annular channel of very small dimensions. The annular channel simulates the flow area around a model of a fuel rod in the VVER nuclear reactor. The annular channel holds spacers which create obstacles to fluid flow. The spacers serve a number of important purposes. In the real nuclear reactor, the spacer holds a fuel rod in the fuel rod bundle. Another important function of the spacer is to influence the flow field characteristics, especially turbulence size, by the shape of the spacer. The value of the turbulence regulates the intensity of heat transfer between the fuel rod and the fluid. Therefore, it is very important to provide a correct description and analysis of the flow field behind the obstacle the spacer generates. The paper further looks into the solution of the same task using numerical simulation. The solution of this task consisted of setting the suitable boundary conditions and of setting the turbulence model for the numerical simulation. The result is a comparison of the flow field characteristics from the experimental measurement and the findings of the numerical simulation. The numerical simulation was carried out using commercial CFD software package, FLUENT.

  10. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  11. CFD Simulation of Thermal-Hydraulic Benchmark V1000CT-2 Using ANSYS CFX

    Directory of Open Access Journals (Sweden)

    Thomas Höhne

    2009-01-01

    Full Text Available Plant measured data from VVER-1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam generator through the steam valves at low reactor power and with all main coolant pumps in operation. CFD calculations have been performed using a numerical grid model of 4.7 million tetrahedral elements. The Best Practice Guidelines in using CFD in nuclear reactor safety applications has been used. Different advanced turbulence models were utilized in the numerical simulation. The results show a clear sector formation of the affected loop at the downcomer, lower plenum and core inlet, which corresponds to the measured values. The maximum local values of the relative temperature rise in the calculation are in the same range of the experiment. Due to this result, it is now possible to improve the mixing models which are usually used in system codes.

  12. Theoretical Study of Steam Condensation Induced Water Hammer Phenomena in Horizontal Pipeline

    CERN Document Server

    Barna, Imre Ferenc

    2014-01-01

    We investigate steam condensation induced water hammer (CIWH) phenomena and present new theoretical results. We use the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. The applied two-fluid model shows some similarities to Relap5 which is widely used in the nuclear industry to simulate nuclear power plant accidents. This model was validated with different CIWH experiments which were performed in the PMK-2 facility, which is a full-pressure thermo-hydraulic model of the nuclear power plant of VVER-440/312 type in the Energy Research Center of the Hungarian Academy of Sciences in Budapest and in the Rosa facility in Japan. In our recent study we show the first part of a planned large database which will give us the upper and lower flooding mass flow ...

  13. Modified Block Newton method for the lambda modes problem

    Energy Technology Data Exchange (ETDEWEB)

    González-Pintor, S., E-mail: segonpin@isirym.upv.es [Departamento de Ingeniería Química y Nuclear, Universidad Politécnica de Valencia, Camino de Vera 14, 46022 Valencia (Spain); Ginestar, D., E-mail: dginestar@mat.upv.es [Instituto de Matemática Multidisciplinar, Universidad Politécnica de Valencia, Camino de Vera 14, 46022 Valencia (Spain); Verdú, G., E-mail: gverdu@iqn.upv.es [Departamento de Ingeniería Química y Nuclear, Universidad Politécnica de Valencia, Camino de Vera 14, 46022 Valencia (Spain)

    2013-06-15

    Highlights: ► The Modal Method is based on expanding the solution in a set of dominant modes. ► Updating the set of dominant modes improve its performance. ► A Modified Block Newton Method, which use previous calculated modes, is proposed. ► The method exhibits a very good local convergence with few iterations. ► Good performance results are also obtained for heavy perturbations. -- Abstract: To study the behaviour of nuclear power reactors it is necessary to solve the time dependent neutron diffusion equation using either a rectangular mesh for PWR and BWR reactors or a hexagonal mesh for VVER reactors. This problem can be solved by means of a modal method, which uses a set of dominant modes to expand the neutron flux. For the transient calculations using the modal method with a moderate number of modes, these modes must be updated each time step to maintain the accuracy of the solution. The updating modes process is also interesting to study perturbed configurations of a reactor. A Modified Block Newton method is studied to update the modes. The performance of the Newton method has been tested for a steady state perturbation analysis of two 2D hexagonal reactors, a perturbed configuration of the IAEA PWR 3D reactor and two configurations associated with a boron dilution transient in a BWR reactor.

  14. NodHex3D: An application for solving the neutron diffusion equations in hexagonal-Z geometry and steady state; NodHex3D: Una aplicacion para solucionar las ecuaciones de difusion de neutrones en geometria hexagonal-Z y estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: jaime.esquivel@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico)

    2014-10-15

    The system called NodHex3D is a graphical application that allows the solution of the neutron diffusion equation. The system considers fuel assemblies of hexagonal cross section. This application arose from the idea of expanding the development of neutron own codes, used primarily for academic purposes. The advantage associated with the use of NodHex3D, is that the kernel configuration and fuel batches is dynamically without affecting directly the base source code of the solution of the neutron diffusion equation. In addition to the kernel configuration to use, specify the values for the cross sections for each batch of fuel used, these values are: diffusion coefficient, removal cross section, absorption cross section, fission cross section and dispersion cross section. Important also, considering that the system is able to perform calculations for various energy groups. As evidence of the operation of NodHex3D, was proposed to model three-dimensional core of a nuclear reactor VVER-1000, based on the reference problem AER-FCM-101. The configuration of the reactor core consists of fuel assemblies (25 batches), composed of seven distinct materials, one of which reflector material, vacuum boundary conditions on the surface delimiting the reactor core. The diffusion equation for two energy groups solves, obtaining the value of the effective neutron multiplication factor. The obtained results are compared to those documented in the reference problem and by 3-DNT codes. (Author)

  15. US-Russian collaboration for enhancing nuclear materials protection, control, and accounting at the Elektrostal uranium fuel-fabrication plant

    Energy Technology Data Exchange (ETDEWEB)

    Smith, H. [Los Alamos National Lab., NM (United States); Allentuck, J. [Brookhaven National Lab., Upton, NY (United States); Barham, M. [Oak Ridge National Lab., TN (United States); Bishop, M. [Sandia National Labs., Albuquerque, NM (United States); Wentz, D. [Lawrence Livermore National Lab., CA (United States); Steele, B.; Bricker, K. [Pacific Northwest National Lab., Richland, WA (United States); Cherry, R. [USDOE, Washington, DC (United States); Snegosky, T. [Dept. of Defense, Washington, DC (United States). Defense Nuclear Agency

    1996-09-01

    In September 1993, an implementing agreement was signed that authorized collaborative projects to enhance Russian national materials control and accounting, physical protection, and regulatory activities, with US assistance funded by the Nunn-Lugar Act. At the first US-Russian technical working group meeting in Moscow in February 1994, it was decided to identify a model facility where materials protection, control, and accounting (MPC and A) and regulatory projects could be carried out using proven technologies and approaches. The low-enriched uranium (LEU or RBMK and VVER) fuel-fabrication process at Elektrostal was selected, and collaborative work began in June 1994. Based on many factors, including initial successes at Elektrostal, the Russians expanded the cooperation by proposing five additional sites for MPC and A development: the Elektrostal medium-enriched uranium (MEU or BN) fuel-fabrication process and additional facilities at Podolsk, Dmitrovgrad, Obninsk, and Mayak. Since that time, multilaboratory teams have been formed to develop and implement MPC and A upgrades at the additional sites, and much new work is underway. This paper summarizes the current status of MPC and A enhancement projects in the LEU fuel-fabrication process and discusses the status of work that addresses similar enhancements in the MEU (BN) fuel processes at Elektrostal, under the recently expanded US-Russian MPC and A cooperation.

  16. Implementation of a fast running full core pin power reconstruction method in DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Torres, Armando Miguel [Instituto Nacional de Investigaciones Nucleares, Department of Nuclear Systems, Carretera Mexico – Toluca s/n, La Marquesa, 52750 Ocoyoacac (Mexico); Sanchez-Espinoza, Victor Hugo, E-mail: victor.sanchez@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-vom-Helmhotz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Kliem, Sören; Gommlich, Andre [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany)

    2014-07-01

    Highlights: • New pin power reconstruction (PPR) method for the nodal diffusion code DYN3D. • Flexible PPR method applicable to a single, a group or to all fuel assemblies (square, hex). • Combination of nodal with pin-wise solutions (non-conform geometry). • PPR capabilities shown for REA of a Minicore (REA) PWR whole core. - Abstract: This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions.

  17. Risk-informed regulation and safety management of nuclear power plants--on the prevention of severe accidents.

    Science.gov (United States)

    Himanen, Risto; Julin, Ari; Jänkälä, Kalle; Holmberg, Jan-Erik; Virolainen, Reino

    2012-11-01

    There are four operating nuclear power plant (NPP) units in Finland. The Teollisuuden Voima (TVO) power company has two 840 MWe BWR units supplied by Asea-Atom at the Olkiluoto site. The Fortum corporation (formerly IVO) has two 500 MWe VVER 440/213 units at the Loviisa site. In addition, a 1600 MWe European Pressurized Water Reactor supplied by AREVA NP (formerly the Framatome ANP--Siemens AG Consortium) is under construction at the Olkiluoto site. Recently, the Finnish Parliament ratified the government Decision in Principle that the utilities' applications to build two new NPP units are in line with the total good of the society. The Finnish utilities, Fenno power company, and TVO company are in progress of qualifying the type of the new nuclear builds. In Finland, risk-informed applications are formally integrated in the regulatory process of NPPs that are already in the early design phase and these are to run through the construction and operation phases all through the entire plant service time. A plant-specific full-scope probabilistic risk assessment (PRA) is required for each NPP. PRAs shall cover internal events, area events (fires, floods), and external events such as harsh weather conditions and seismic events in all operating modes. Special attention is devoted to the use of various risk-informed PRA applications in the licensing of Olkiluoto 3 NPP. © 2012 Society for Risk Analysis.

  18. CATHARE2 calculation of SPE-3 test small break loca on PMK facility

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, E.; Radet, J. [Institut de Protection et de Surete Nucleaire, Cadarache (France)

    1995-09-01

    Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a {open_quotes}bleed and feed{close_quotes} procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation.

  19. Scientific basis for modelling and calculation of acoustic vibrations in the nuclear power plant coolant

    Science.gov (United States)

    Proskuryakov, K. N.

    2017-11-01

    Created new scientific direction: “Diagnosis, prognosis and prevention of vibration - acoustic resonances in the nuclear power plant (NPP) equipment. The possibility of using methods for calculating and analyzing electric oscillation systems in the study of the properties of acoustic systems with a two-phase medium is proved, based on the similarity of the differential equations describing the state of these systems. Is shown that the developed methods can be used to predict and prevent the occurrence of vibration - acoustic resonances in the NPP equipment. Is shown that the volume of pressurizer at NPPs with VVER and PWR as well as boiling water reactor that exploded at Japan’s NPP Fukushima Daiichi is a Helmholtz resonator, which contain water and steam volumes and able many times increases the impact on them of outside periodic oscillations. Paper presents most important results published long before the severe accidents at NPPs Three Mile Island (TMI), Chernobyl and Fukushima Daiichi that could be used for the prediction of a severe accident scenario, identification of measuring data and process control in order to minimize the damage. Worked out results also could be useful in another industrial technologies based on applications of single and two-phase flows.

  20. NONUNIFORMITIES OF TWO-PHASE COOLANT DISTRIBUTION IN A HEAT GENERATING PARTICLES BED

    Directory of Open Access Journals (Sweden)

    V. V. Sorokin

    2014-01-01

    Full Text Available Sufficient atomic power generation safety increase may be done with microfuel adapting to reactor plants with water coolant. Microfuel particle is a millimeter size grain containing fission material core in a protecting coverage. The coverage protects fuel contact with coolant and provides isolation of fission products inside. Well thermophysical properties of microfuel bed in a direct contact with water coolant excludes fuel overheating when accidents. Microfuel use was suggested for a VVER, а direct flow reactor for superheat steam generation, a reactor with neutron spectra adjustment by the steam partial content varying in the coolant.Nonuniformities of two-phase coolant distribution in a heat generating particles bed are predicted by calculations in this text. The one is due to multiple-valuedness of pressure drop across the bed on the steam quality dependency. The nonuniformity decreases with flow rate and particle size growths absolute pressure diminishing while porosity effect is weak. The worse case is for pressure quality of order of one. Some pure steam filled pores appears parallel to steam water mixture filled pores, latter steam quality is less than the mean of the bed. Considering this regime for the direct flow reactor for superheat steam generation we predict some water drops at the exit flow. The two-phase coolant filtration with subcooled water feed is unstable to strong disturbance effects are found. Uniformity of two-phase coolant distribution is worse than for one-phase in the same radial type reactor.

  1. Development of the core safety regulation technology for the SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Zin; Kim, Do Sam; Lee, Kyeong Taek; Park, Young Ryoung; Lee, Gil Soo; Kim, Jong Woon; Yun, Sung Hwan; Lee, Jae Jun; Lee, Myung Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2003-06-15

    As the SMART-P is different from existing general reactors, new regulation technology is required to understand and assess the SMART-P for its regulatory reviews. One of the these technologies is related to the core design analysis. Because the SMART-P used metallic fuels, this study also collects general metallic nuclear fuel data and SMART-P's metallic fuel data from the materials studied by KAERI. The core design methodologies of KWU, ABB-CE, Westinghouse, Studsvik, Scandpower, US NRC and domestic research centers were investigated. Specially, The Hellios lattice core was studied for hexagonal nuclear fuel assembly calculation. Also, the VVER-1000 benchmark problem was analyzed by the PARCS code which has been developed by U.S. NRC. In this study, a AFEN-based computing code KORDAX os developed for the regulatory review of the SMART-P. KORDAX which is a nodal code using AFEN method dose not use transverse integration and this it can give higher accuracy results. Also, Because KORDAX is useful for hexagonal core and uses a method different with the core design code of the SMART-P developed by KAERI, it is judged that KORDAX can be an independent and reliable regulation verification code. In the next year study, HELIOS will be further studied as a core lattice code, and a hexagonal kinetics code which is based on AFEN method will be developed more systematically.

  2. Simulation of the Thermal Hydraulic Processes in the Horizontal Steam Generator with the Use of the Different Interfacial Friction Correlations

    Directory of Open Access Journals (Sweden)

    Vladimir Melikhov

    2011-01-01

    Full Text Available The horizontal steam generator (SG is one of specific features of Russian-type pressurized water reactors (VVERs. The main advantages of horizontal steam generator are connected with low steam loads on evaporation surface, simple separation scheme and high circulation ratio. The complex three-dimensional steam-water flows in the steam generator vessel influence significantly the processes of the steam separation, distribution, and deposition of the soluble and nonsoluble impurities and determine the efficiency and reliability of the steam generator operation. The 3D code for simulation of the three-dimensional steam-water flows in the steam generator could be effective tool for design and optimization of the horizontal steam generator. The results of the code calculations are determined mainly by the set of the correlations describing interaction of the steam-water mixture with the inner constructions of the SG and interfacial friction. The results obtained by 3D code STEG with the usage of the different interfacial friction correlations are presented and discussed in the paper. These results are compared with the experimental ones obtained at the experimental test facility PGV-1500 constructed for investigation of the processes in the horizontal steam generator.

  3. Effect of fission fragment on thermal conductivity via electrons with an energy about 0.5 MeV in fuel rod gap

    Directory of Open Access Journals (Sweden)

    F Golian

    2017-02-01

    Full Text Available The heat transfer process from pellet to coolant is one of the important issues in nuclear reactor. In this regard, the fuel to clad gap and its physical and chemical properties are effective factors on heat transfer in nuclear fuel rod discussion. So, the energy distribution function of electrons with an energy about 0.5 MeV in fuel rod gap in Busherhr’s VVER-1000 nuclear reactor was investigated in this paper. Also, the effect of fission fragments such as Krypton, Bromine, Xenon, Rubidium and Cesium on the electron energy distribution function as well as the heat conduction via electrons in the fuel rod gap have been studied. For this purpose, the Fokker- Planck equation governing the stochastic behavior of electrons in absorbing gap element has been applied in order to obtain the energy distribution function of electrons. This equation was solved via Runge-Kutta numerical method. On the other hand, the electron energy distribution function was determined by using Monte Carlo GEANT4 code. It was concluded that these fission fragments have virtually insignificant effect on energy distribution of electrons and therefore, on thermal conductivity via electrons in the fuel to clad gap. It is worth noting that this result is consistent with the results of other experiments. Also, it is shown that electron relaxation in gap leads to decrease in thermal conductivity via electrons

  4. Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach*

    Directory of Open Access Journals (Sweden)

    Bury Tomasz

    2015-06-01

    Full Text Available The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

  5. Investigation of two-phase flow structure in model of draught pipe of water boiling reactor VK-300

    Energy Technology Data Exchange (ETDEWEB)

    Efanov, A.D.; Kuznetzov, Y.N.; Kaliakin, S.G.; Lisitza, F.D.; Remizov, O.V.; Serdun, N.P. [Institute of Physics and Power Engineering State Scientific Center, Obninsk (Russian Federation)

    2001-07-01

    VK-300 reactor represents a vessel-type boiling reactor with integral arrangement of assemblies and in-vessel steam separation at one-circuit scheme. The circuit consists of core, draught pipes, and separation facilities. The vessel of VK-300 reactor is chosen on the base of the dimensions of that of VVER-1000 reactor. The following thermal-hydraulic parameters of nuclear power plant (NPP) were investigated experimentally: dependence of void fraction upon the steam quality in mixing chamber (on the draught section input); pressure losses at different, specific zones of up-flow and down-flow sections of the circuit with free circulation; degree of steam separation in the separating chamber (at the first step of phase separation) and its dependence upon steam quality; structure of steam-water flow in draught pipes (distribution of phases over the draught pipe cross- section); presence of steam hovering and height of this hovering in inter-pipe space of draught section. (author)

  6. Passive Core Cooling Systems for Next Generation NPPs: Characteristics and State of the Art

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, Andrey; Soshkina, Alexandra [Institute for Physics and Power Engineering by A.I. Leypunsky, 1 Bondarenko sq. Obninsk, 249033 (Russian Federation)

    2008-07-01

    Among nuclear power generation plants, light water reactors are mainly used at present, and are anticipated to be predominant in the future. To improve the light water reactors the development of the LWRs for the next generation is carried out at various organizations. For example, in the USA the Westinghouse AP-1000 is based on proven technology but with an emphasis on passive safety features. The reactor passive core cooling systems include the core makeup tanks system, passive residual heat removal heat exchanger and in-containment refuelling water storage tank. In Russia has been developed the so-called NPP-2006 project of a VVER-1200 nuclear power plant with a V-392M reactor unit. To provide the safety, protection passive systems which do not depend upon human errors are widely used in this project. Among these are hydro-tanks of the second stage and passive heat removal system. In the presented paper an overview of passive core cooling systems for next generation NPPs is given. (authors)

  7. Competence and experience for commercial demolition of nuclear research facilities; Kompetenz und Erfahrung fuer den wirtschaftlich orientierten Rueckbau kerntechnischer Forschungsanlagen

    Energy Technology Data Exchange (ETDEWEB)

    Starke, Holger [Babcock Noell GmbH, Wuerzburg (Germany)

    2010-10-15

    By international comparison, the demolition of nuclear facilities in Germany began early, i.e. in the early 1980s. Those projects constituted virgin territory in the field of nuclear technology. There was no experience in applying existing codes, rules and regulations to the dismantling of activated and contaminated structures so as to protect personnel, the environment, and the public. Based on experience accumulated in the demolition of commercially used plants, Babcock Noell GmbH (BNG) handled some first projects in German research installations. This experience then allowed the company to solicit other demolition projects in research installations in other European countries. One of the advantages which turned out to be useful was BNG's experience in the Russian VVER nuclear power plant line (water-water reactor, Russian research reactor line) plus the fact that several research reactors of that design were to be decommissioned and demolished in countries in Eastern Europe. The objectives, organization and implementation of demolition projects of nuclear research installations are outlined for these facilities: - Rossendorf research reactor (RFR), Dresden-Rossendorf, Germany; - Joint Research Centre (JRC), Ispra, Italy, with 3 research reactors, various laboratories and waste stores; - research reactor of the Salaspils, Latvia, Research Center; - the FMRB reactor of the Federal Institute of Physics and Metrology (PTB), Brunswick, Germany; - the FRF research reactor, Frankfurt, Germany and - demolition of the Magurele, Romania, research reactor. (orig.)

  8. Containment PT Analysis in case of installation of PCCS using CAP Code

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yeon Jun; Hong, Soon Joon [Future and Challenge Technology Co., Yongin (Korea, Republic of); Kim, Gon Han; Cheon, Jong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The goal of this project is developing of the conceptual design of containment passive cooling system, including PMCCS (Passive Molten Core Cooling System) and furthermore, applying to APR+ and iPOWER reactor finally. Advanced researches on PCCS were carried out by global nuclear industrial companies and the several types of reactor, such as AP1000, ESBWR and VVER, were introduced. Using the different way, these reactor types, however, are devised to mitigate the consequence of LOCA accident and remove the long-term decay heat. To confirm the reliability of PCCS function during an accident progress, experimental proofs and computational evaluations are vital. To evaluate the PCCS performance, CAP code simulation is preliminarily conducted on the ShinKori 3/4 DEDLSB accident. Two condensation models, Uchida and Dehbi's correlation, are tested for the condensation model applied on PCCS condensation tube outside surface. As compared with the existing spray system, it revealed the good performance in terms of containment pressure reduction. On the other hand, re-pressurization with the start of PCCT coolant temperature increment is observed also.

  9. Effects of B4C control rod degradation under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Si-Won; Park, Sang-Gil; Han, Sang-Ku [Atomic Creative Technology Co., Daejeon (Korea, Republic of)

    2016-10-15

    Boron carbide (B4C) is widely used as absorber material in western boiling water reactor (BWR), some PWR, EPR and Russian RBMK and VVERs. B4C oxidation is one of the important phenomena of in-vessel. In the present paper, the main results and knowledge gained regarding the B4C control rod degradation from above mentioned experiments are reviewed and arranged to inform its significance on the severe accident consequences. In this paper, the role of B4C control rod oxidation and the subsequent degradation on the severe accident consequences is reviewed with available literature and report of previous experimental program regarding the B4C oxidation. From this review, it seems that the contribution of this B4C oxidation on the accident progression to the further severe accident situation is not negligible. For the future work, the extensive experimental data interpretation will be performed to assess quantitatively the effect of B4C oxidation and degradation on the various postulated severe accident conditions.

  10. Two decades of experience with more than 750 CASTOR and CONSTOR transport and storage casks

    Energy Technology Data Exchange (ETDEWEB)

    Kuehne, B. [GNB Gesellschaft fuer Nuklear-Behaelter mbH, Essen (Germany); Kuehl, H. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2004-07-01

    In 1983 the world-wide first dual purpose transport and storage cask - a CASTOR {sup registered} Ic-DIORIT - was loaded in Wuerenlingen/ Switzerland. Meanwhile CASTOR {sup registered} casks are used at 24 sites on four continents. Spent fuel assemblies of PWR, BWR, VVER, RBMK, FBR, MTR and THTR as well as vitrified high active waste canisters are transported and/or stored in these kinds of monolithic metal casks. MOX spent fuel of PWR and BWR has been loaded, too. Starting in the mid of the 90s, GNB developed the new CONSTOR {sup registered} cask concept, which is based on a double liner technology with a layer of heavy concrete as shielding material inbetween. This CONSTOR {sup registered} cask concept fulfils all design criteria for transport and for storage given by the IAEA recommendations and by national authorities. Up to now, more than 750 CASTOR {sup registered} and CONSTOR {sup registered} casks have been used for transports or/and loaded for longterm interim storage. More than two decades of storage experience attest to the excellent behavior of the casks including the metallic gaskets and the tightness monitoring system. Detailed measurements of temperatures and of gamma and neutron dose rates have shown in each case that the safety requirements have been fulfilled. These measurements allowed to reduce unnecessary safety margins to optimize the benefit for the user.

  11. Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1997-12-31

    The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place. 6 refs.

  12. Replacement of nickel sealing rings by expanded graphite sealing rings -upgrading of SG primary collector flange connection

    Energy Technology Data Exchange (ETDEWEB)

    Cikryt, F.; Bednarek, L.; Kusyn, L. [Vitkovice, Ostrava (Switzerland)

    1997-12-31

    One of the most loaded parts of a steam generator of VVER 440 MW type are the bolts and thread holes of the primary collector cover sealing set. The strength calculations and tensometric measurings performed during operation proved the high degree of a load on the bolts. The conditions of the stress limitation are not met in some cases according to the pertinent standards. The untightnesses at nickel rings occurred during putting the units of Jaslovske Bohunice and Dukovany nuclear power stations into operation. With regard to improve the reliability, the producer has taken measures to improve the quality of the rings and users have introduced more strict regulations for bolts tightening. Due to these measures the high reliability of the set has been obtained from point of view of the tightness, but substantial reduction of bolts and holes threads loading have not been obtained. Several years operation experience proved relatively low service of bolts, damage of thread holes and sealing grooves. As the degree of mechanical load is one of the vital parameters influencing the damage of sealing set, in 1996 we started with the works aimed at a possibility of nickel sealing rings replacement for a more modern type of sealing which assure the higher reliability and service life of the individual part of sealing set under the reduced load.

  13. Criticality and shielding study for reracking of the spent fuel storage of NPP Jaragua; Estudio de la criticalidad y el blindaje del almacen de combustible irradiado de la Central Electronuclear de Juragua para redes compactas

    Energy Technology Data Exchange (ETDEWEB)

    Guerra Valdes, Ramiro; Lopez Aldama, Daniel; Rodriguez Gual, Maritza; Garcia Yip, Fernando; Alvarez Cardona, Caridad [Centro de Tecnologia Nuclear, La Habana (Cuba)

    1996-07-01

    Annually one third of the fuel assemblies are discharged from VVER-440 reactor core. After 2 years of decay in the refueling pool, these assemblies are transferred to the spent fuel storage pool. With two units in operation, it would exhaust its storage capability in about 10 years. According to the trend of extending the interim storage period, the reracking of the spent fuel storage pool has become a viable option to enlarge storage capability at the nuclear power plant. The present paper deals with the criticality and shielding analysis for reracking of the spent fuel storage pool of NPP Juragua. The WIMS/D4 lattice code is used for the criticality study. For doses calculations, the source strength is estimated with ORIGEN 2 and the shielding problem is solved with the combination of the code ANISN and the multigroup library CASK. It is shown that it is possible to compact the storage rack in a factor of 1.98 using 3 mm thick boron steel clads. While the source is nearly doubled in the pool, the doses in its boundaries are increased approximately in only 1.24 times. (author)

  14. Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

    Directory of Open Access Journals (Sweden)

    E. A. Kuleshova

    2017-01-01

    Full Text Available This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C. The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

  15. Comprehensive investigation of the corrosion state of the heat exchanger tubes of steam generators. Part II. Chemical composition and structure of tube surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Homonnay, Z. [Department of Nuclear Chemistry, Faculty of Science, Eoetvoes University, H-1518 Budapest, P.O. Box 32 (Hungary)]. E-mail: homonnay@ludens.elte.hu; Kuzmann, E. [Research Group for Nuclear Methods in Structural Chemistry, Hungarian Academy of Sciences, Eoetvoes University, Budapest (Hungary); Varga, K. [Department of Radiochemistry, University of Veszprem, H-8201 Veszprem, PO Box: 158 (Hungary)]. E-mail: vargakl@almos.vein.hu; Nemeth, Z. [Department of Radiochemistry, University of Veszprem, H-8201 Veszprem, PO Box: 158 (Hungary); Szabo, A. [Department of Radiochemistry, University of Veszprem, H-8201 Veszprem, PO Box: 158 (Hungary); Rado, K. [Department of Radiochemistry, University of Veszprem, H-8201 Veszprem, PO Box: 158 (Hungary); Mako, K.E. [Department of Silicate and Materials Engineering, University of Veszprem, Veszprem (Hungary); Koever, L. [Section of Electron Spectroscopy, Institute of Nuclear Research, H-4001 Debrecen (Hungary); Cserny, I. [Section of Electron Spectroscopy, Institute of Nuclear Research, H-4001 Debrecen (Hungary); Varga, D. [Section of Electron Spectroscopy, Institute of Nuclear Research, H-4001 Debrecen (Hungary); Toth, J. [Section of Electron Spectroscopy, Institute of Nuclear Research, H-4001 Debrecen (Hungary); Schunk, J. [Paks NPP, Paks (Hungary); Tilky, P. [Paks NPP, Paks (Hungary); Patek, G. [Paks NPP, Paks (Hungary)

    2006-01-01

    In the frame of a project dealing with the comprehensive study of the corrosion state of the steam generators of the Paks Nuclear Power Plant, Hungary, surface properties (chemical and phase compositions) of the heat exchanger tubes supplied by the power plant were studied by Moessbauer spectroscopy (CEMS), X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) methods. The work presented in this series provides evidence that chemical decontamination of the steam generators by the AP-CITROX technology does exert a detrimental effect on the chemical composition and structure of the protective oxide film grown-on the inner surfaces of heat exchanger piping. As an undesired consequence of the decontamination technology, a 'hybrid' structure of the amorphous and crystalline phases is formed in the outermost surface region (within a range of 11 {mu}m). The constituents of this 'hybrid' structure exhibit great mobility into the primary coolant under normal operation of the VVER type reactor.

  16. ATHLET. Mod 3.0 Cycle A. Validation

    Energy Technology Data Exchange (ETDEWEB)

    Lerchl, G.; Austregesilo, H.; Glaeser, H.; Hrubisko, M.; Luther, W.

    2012-09-15

    ATHLET is an advanced best-estimate code which has been initially developed for the simulation of design basis and beyond design basis accidents (without core degradation) in light water reactors, including VVER and RBMK reactors. Furthermore, this program version enables the simulation of further working fluids like helium and liquid metals. The one-dimensional, two-phase fluiddynamic models are based on a five-equation model supplemented by a full-range drift-flux model, including a dynamic mixture-level tracking capability. Moreover, a two-fluid model based on six conservation equations is provided. The heat conduction and heat transfer module allows a flexible simulation of fuel rods and structures. The nuclear heat generation is calculated by a point-kinetics or by a one-dimensional kinetics model. A general control simulation module is provided for a flexible modelling of BOP- and auxiliary plant systems. Systematic code validation is performed by GRS and independent organizations. This Validation Manual is the fourth volume of the ATHLET Code Documentation comprising four volumes. This manual presents an overview about the complete ATHLET validation effort spent up to now. In addition, the results of five test cases simulated with the present ATHLET program version are compared with the experimental data.

  17. INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-09-01

    This is the final report on the INSP project entitled, ``Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install and make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant.

  18. V1000CT-1 benchmark analyses with the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems

    Energy Technology Data Exchange (ETDEWEB)

    Yaroslav Kozmenkov; Ulrich Grundmann; Soeren Kliem; Ulrich Rohde; Frank-Peter Weiss [Forschungszentrum Rossendorf, Institut fuer Sicherheitsforschung Postfach 510119, D 01314 Dresden (Germany)

    2005-07-01

    Full text of publication follows:Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and 3 of 4 MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Control rods were not changing their positions during the transient. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical nodalization schemes, MCP characteristics, boundary conditions and the benchmark-specified nuclear data library. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermohydraulic models of the system codes RELAP5 and ATHLET. (authors)

  19. Institute for Energy Technology -Annual report 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    Research at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The main nuclear program is centered on the OECD Halden Reactor Project. 19 participating countries and about 100 organisations is involved in the project. The Project is operated by a staff of 280 persons. In the autumn of 1996 the participating organizations reached agreement to continue their research collaboration for a further 3-year period (1997 to 1999). An extensive experimental program was carried out in 1996 using the Halden reactor (HBWR), partly for the joint international program, and partly for contract work for member countries. The main aim of this work is to improve the safety and reliability of existing nuclear power plants. The experimental equipment in the Halden reactor makes it ideal for simulating various operating conditions in different types of rectors. Processes such as corrosion in fuel cladding materials and fracture propagation in irradiated materials under the influence of additives in the coolant water can be studied. In an on-going study, fuel of Russian origin is being compared with modern western fuel. The results, being the first of their kind that are openly available, form an important bases for safety assessments of Russian VVER reactors. The man-machine laboratory is used to study how new technologies influence the operator and to develop computer based systems for improving the safety and accessibility of complex processes.

  20. Mechanizmy patogenetyczne IgE-zależne i IgE-niezależne w nadwrażliwości pokarmowej u dzieci i młodzieży

    Directory of Open Access Journals (Sweden)

    Maciej Kaczmarski

    2010-12-01

    Full Text Available Według aktualnie obowiązującej terminologii Europejskiej Akademii Alergologii i Immunologii Klinicznej nadwrażliwość organizmu to nieprawidłowa, opaczna i powtarzająca się reakcja na spożyty lub spożywany pokarm, który jest dobrze tolerowany przez osoby zdrowe. Ze względu na mechanizm patogenetyczny dzielimy ją na immunologiczną lub nieimmunologiczną. Do pierwszej grupy należą chorzy z alergią pokarmową, u których objawy chorobowe są wynikiem udziału przeciwciał IgE. Alergicznie na szkodliwy pokarm reaguje około 8% małych dzieci, w większości nadwrażliwość pokarmowa jest u nich wywołana reakcją organizmu na białka mleka krowiego. Za najważniejsze alergeny odpowiedzialne za proces uczulenia i/lub rozwoju nadwrażliwości pokarmowej uważane są: białka mleka krowiego, białka jaja, ryby, skorupiaki, mięczaki, orzechy, soja, pszenica. Klinika alergii pokarmowej jest bardzo bogata. Rozpoznanie nadwrażliwości pokarmowej to trudny i wieloetapowy proces, gdyż nie dysponujemy łatwym, tanim, czułym i swoistym testem ułatwiającym to zadanie. Podstawą rozpoznania alergii jest udowodnienie zależności przyczynowo-skutkowej między spożyciem szkodliwego pokarmu a wystąpieniem objawów. Z metod alergologiczno-immunologicznych dysponujemy możliwością rozpoznawania alergii pokarmowej IgE-zależ- nej i oznaczenia w surowicy IgE całkowitej oraz alergenowoswoistych IgE. Ten mechanizm patogenetyczny reakcji wykrywa się poprzez wykonanie testów skórnych typu prick z alergenami pokarmowymi i/lub powietrznopochodnymi. W odniesieniu do reakcji klinicznych wywołanych mechanizmami IgE-niezależnymi dysponujemy możliwością wykonania płatkowych testów skórnych z alergenami pokarmowymi. W nadwrażliwości pokarmowej nie ma jednego, uniwersalnego testu diagnostycznego. Dla postawienia właściwej diagnozy decydujące znaczenie ma doustna próba prowokacji i eliminacji. Podstawą w leczeniu przyczynowym tej

  1. Układ zagrody holenderskiej na przykładzie środkowego Mazowsza

    Directory of Open Access Journals (Sweden)

    Ewa Mariola Zaraś-Januszkiewicz

    2014-04-01

    Full Text Available Ważnymi elementami współczesnego krajobrazu kulturowego środkowej części Mazowsza, określanej w historycznym ujęciu jako ziemia warszawska, są pozostałości po osadnictwie holenderskim. Na szczególną uwagę zasługuję liniowy sposób rozmieszczenia domów mieszkalnych i zabudowań gospodarczych, a także ogrodów przydomowych w obrębie pojedynczej zagrody oraz dróg łączących poszczególne obejścia. Zarówno zabudowania, jak i drogi łączące obejścia sąsiadujących ze sobą gospodarzy lokalizowano na wyniesieniach terenu – naturalnych lub częściej sztucznych (zwanych terpami i trytfami. W strefie wejściowej zagrody znajdował się najczęściej ogród przydomowy i sad, za którym wznosił się z reguły jeden obszerny dom z liniowo uszeregowanymi pomieszczeniami mieszkalnymi, następnie spichlerzem i na końcu pomieszczeniami dla zwierząt (dom typu langhoff. Do dziś struktury te są dobrze widoczne w przypadku osad zakładanych rzędowo w pobliżu rzek. Innym elementem typowym dla osadnictwa holenderskiego było stosowanie nasadzeń z drzew jako elementów przeciwdziałających zniszczeniom powodziowym. W przypadku Mazowsza do dziś bardzo dobrze widoczne są rzędowe nasadzenia ogławianych wierzb (Salix alba i Salix fragilis, rzadziej topól (Populus alba. Nasadzenia te towarzyszyły drogom, miedzom, rowom odwadniającym. Wiele z tych elementów i struktur przetrwało do dziś jedynie w szczątkowej postaci.

  2. Rys historyczny i analiza dendrologiczna zespołu pałacowego Podzamcze w Bychawie

    Directory of Open Access Journals (Sweden)

    Marek Dąbski

    2016-06-01

    Full Text Available Zespół pałacowy Podzamcze w Bychawie, z ruinami zamku, drewnianym dworem i budynkami gospodarczymi to jeden z ciekawszych zabytków regionu bychawskiego, wpisany na listę zabytków woj. lubelskiego. Budowę bychawskiego zamku rozpoczął Mikołaj Pilecki herbu Leliwa w 1539 roku. Zamek położony był na półwyspie, otoczony z trzech stron stawami, odcięty od lądu fosą i połączony z podzamczem mostem zwodzonym. Bychawa często zmieniała swoich właścicieli – byli nimi m.in. ród Myszkowskich z Pińczowa, Andrzej Leopold Ossowski, Dominik Stoiński, Karol Scipio del Campo, rodzina Łaniewskich. W XIX w. otoczenie pałacu obsadzono topolami włoskimi, cyprysami i żywotnikami. Pomiędzy pałacem a stawem znajdowały się tarasowe ogrody włoskie, a w dalszej części terenu rozciągał się park typu krajobrazowego, wzdłuż którego dróg poprowadzono szpalery i aleje drzew. Pałac opustoszał po pożarze, który miał miejsce w latach 80. XIX w. Obecnie stanowi malowniczą ruinę, niepozbawioną wartości historycznych. Drzewostan parkowy reprezentowany jest przez 21 taksonów, a największa liczba egzemplarzy należy do gatunków: Acer platanoides L. (89 szt., Fraxinus excelsior L. (80, Tilia cordata Mill. (54 i Populus alba L. (38. Wśród podszytu najliczniejsze grupy stanowią krzewy z gatunków: Sambucus nigra L., Cornus mas L., Salix purpurea L. i Corylus avellena L. Projekt rewaloryzacji obiektu powinien zmierzać do poprawy walorów estetycznych obszaru, ekspozycji wzgórza zamkowego z odtworzeniem zatartych osi widokowych, podkreślenia kulturowej i dydaktycznej funkcji miejsca oraz stworzenie dla mieszkańców przyjaznego i dostępnego terenu rekreacji.

  3. BADANIA KORELACJI PIERWIASTKÓW ŚLADOWYCH W ŚRODOWISKU GLEBOWO – ROŚLINNYM PRZY ZASTOSOWANIU METOD GIS

    Directory of Open Access Journals (Sweden)

    Agnieszka PĘKALA

    2016-03-01

    Full Text Available Celem prezentowanego opracowania jest identyfikacja anomalnych koncentracji geochemicznych w glebach oraz korzeniu marchwi pochodzących z ogródków działkowych miasta Przemyśla. Przy zastosowaniu technik kartograficznych wykonano mapy monitoringu środowiska zanieczyszczeń, obejmujących rejestracje miejsc zakładów przemysłowych, wprowadzających wyznaczone pierwiastki śladowe do środowiska przyrodniczego. Technologie te umożliwiły również wizualizację interakcji zachodzących między systemem przyrodniczym typu gleba-roślina i wykonaniu map korelacji występującej w tym układzie. W ramach badań chemicznych, przy zastosowaniu atomowej spektroskopii absorpcyjnej (ASA, wyznaczono koncentrację kadmu, ołowiu i miedzi w roślinach oraz glebach. Badania mineralogiczne przy zastosowaniu dyfraktometrii rentgenowskiej oraz mikroskopii skenningowej, umożliwiły określenie składu fazowego badanych gleb. Wszystkie operacje i czynności związane z analizą przestrzenną i kartowaniem wykonane zostały w opensource`owym systemie QGIS/GRASS. Uzyskane wyniki badań chemicznych dla materiału roślinnego oraz gleb pozwalają stwierdzić, że we wszystkich 14 miejscach pomiarowych została przekroczona górna dopuszczalna granica zawartości Cd podawana w rozporządzeniu 420/211 komisji UE z 2011 roku. Zawartość Cu i Pb we wszystkich próbkach nie przekraczają dopuszczalnych norm. Mapy korelacji pomiędzy środowiskiem glebowym a roślinnym potwierdzają wyniki badań geochemicznych. Miejsca największej koncentracji kadmu pokrywają się z podwyższoną zawartością tego pierwiastka w roślinach.

  4. SYNTEZA ZEOLITÓW DO ADSORPCJI ACETONU

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    Magdalena WARZYBOK

    2016-03-01

    Full Text Available Celem prowadzonych badań było uzyskanie efektywnych adsorbentów do usuwania acetonu z gazów odlotowych. Wybrano zeolit typu Y ze względu na jego wysoką termo stabilność umożliwiającą desorpcję zaadsorbowanego acetonu w temperaturach dochodzących do 1000°C. Proces syntezy składał się z czterech etapów: (1 aktywacji termicznej surowego materiału, (2 starzenia mieszanin reakcyjnych w temperaturze otoczenia, (3 krystalizacji (wysokotemperaturowego ogrzewania składników oraz (4 przemywania i suszenia produktu. Jako substraty wykorzystano naturalne materiały ilaste – bentonit (B, haloizyt (H oraz kaolin (K. Aby zapewnić odpowiedni stosunek molowy Na2O:SiO2:Al2O3:H2O do syntezy zastosowano również krzemionkę (SiO2 oraz roztwory chlorku (NaCl oraz wodorotlenku sodu (NaOH. Podczas badań optymalizowano: temperaturę aktywacji, czas starzenia oraz czas i temperaturę etapu krystalizacji. Wpływ optymalizowanych parametrów na właściwości otrzymanych adsorbentów oceniano na podstawie: masy uzyskanego produktu, straty po prażeniu (LOI [%], stężenia jonów Na+ w przesączach poreakcyjnych (CNa [mg/l] oraz pojemności adsorpcyjnej względem acetonu (qe [mg/g]. Optymalna temperatura aktywacji wyjściowych materiałów ilastych wynosi 600°C. Podnoszenie temperatury aktywacji o kolejne 100°C skutkowało pogarszaniem właściwości adsorbentów. Wydłużenie czasu starzenia i krystalizacji, jak również podwyższenie temperatury etapu krystalizacji poprawia właściwości adsorpcyjne otrzymanych zeolitów. Optymalny czas etapu krystalizacji zależy od rodzaju materiału wyjściowego użytego do syntezy, temperatury aktywacji oraz czasu starzenia. Wydłużenie czasu starzenia mieszanin reakcyjnych pozwala na skrócenie czasu krystalizacji.

  5. „Polska Bibliografia Bibliologiczna” jako źródło danych do badań bibliometrycznych, na przykładzie problematyki bibliotek toruńskich

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    Małgorzata Kowalska

    2011-12-01

    Full Text Available Mimo zmian zachodzących w metodach przekazywania informacji, publikacje drukowane są nadal istotnym środkiem rozpowszechniania wiedzy i posiadają ugruntowaną pozycję w środowisku akademickim. Są one również jednym ze znaczących elementów wartościowania dorobku naukowego poszczególnych pracowników nauki i oceny działalności instytucji badawczych. Coraz częściej do tego celu, obok różnych mierników jakości (impact factor, indeksy Hirscha i Egghe, punktacja Ministerstwa Nauki i Szkolnictwa Wyższego, wykorzystuje się opisy bibliograficzne lub ich wybrane cechy. Na tej podstawie wskazuje się  relacje pomiędzy  dokumentami  oraz  ustala  parametry charakteryzujące ich zbiory i właściwości. Niniejszy artykuł jest próbą wskazania tego typu zależności. Prezentuje on wyniki analizy dorobku dokumentalnego tematycznie dedykowanego bibliotekom w Toruniu, a zarejestrowanego w „Polskiej Bibliografii Bibliologicznej” w latach 1995–2011. Materiał badawczy obejmuje zbiór 264 pozycji bibliograficznych, na który składają się różne typy dokumentów: książki i ich redakcje, wydawnictwa periodyczne, artykuły z czasopism oraz fragmenty prac zbiorowych. Autorka dokonuje analizy ilościowej i treściowej zebranego materiału, zwracając szczególną uwagę zarówno na dynamikę przyrostu publikacji, różnorodność form wydawniczych oraz zakresy tematyczne publikowanych prac, jak i specyfikę „Polskiej Bibliografii Bibliologicznej” jako źródła danych do badań bibliometrycznych.

  6. Ropień tkanek miękkich podudzia wywołany przez Salmonella sp. u pacjentki z reumatoidalnym zapaleniem stawów leczonej adalimumabem

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    Magdalena Marek

    2011-04-01

    Full Text Available Stosowanie leków biologicznych, do których zalicza się m.in. przeciwciałaprzeciwko czynnikowi martwicy nowotworów α, może byćzwiązane ze zwiększoną zapadalnością na zakażenia o etiologiibakteryjnej, wirusowej czy grzybiczej.Supresja układu odpornościowego czasami zmienia typowy przebiegkliniczny infekcji w porównaniu z przebiegiem u osób immunokompetentnych.Obserwuje się tendencję do uogólnienia zakażeniai/lub nietypowego, skąpoobjawowego przebiegu.Przedstawiono przypadek 60-letniej kobiety chorej na reumatoidalnezapalenie stawów o agresywnym przebiegu od ok. 12 lat,obciążonej wieloletnią cukrzycą typu 2, leczonej adalimumabem,metotreksatem i prednizonem, przyjętej do szpitala z powodubolesnej zmiany o charakterze naciekowo-zapalnym na bocznejpowierzchni prawego podudzia. Przeprowadzona diagnostykadoprowadziła do rozpoznania ropnia tkanek miękkich prawegopodudzia o etiologii Salmonella sp. (ryc. 1, 2.Etiologia została potwierdzona badaniem bakteriologicznymwydzieliny uzyskanej po chirurgicznym nacięciu ropnia. Zastosowanoantybiotykoterapię zgodną z antybiogramem oraz leczeniemiejscowe (miejscowo działające leki odkażające i regularnawy mia na sączków, uzyskując bardzo dobry efekt kliniczny (ryc. 3.W cza sie diagnostyki i leczenia odstawiono metotreksat i adali -mumab.Powstanie ropnia było prawdopodobnie poprzedzone skąpoobjawowym,ograniczonym jedynie do stanów gorączkowych, zakażeniemSalmonella i bakteriemią. Należy również wziąć pod uwagęmożliwość nosicielstwa i reaktywacji zakażenia w wyniku stosowanychleków immunosupresyjnych. Z powodu wciąż rosnącej liczby chorych stosujących leki biologicznewarto zwrócić uwagę na to, że zakażenia w tej grupie chorychmogą mieć bardzo nietypowy przebieg oraz etiologię.Decyzja o ponownym stosowaniu leku z grupy leków modyfikującychprzebieg choroby, ewentualnie o ponownej kwalifikacji doleczenia biologicznego, będzie podjęta po zako

  7. Sportovní slavnosti [Sports festivities

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    Anna Hogenová

    2008-02-01

    Full Text Available Článek se soustřeďuje na fenomén slavnosti, jenž je inherentí každému sportovnímu agonu. Slavnost je analyzována jako fenomén setkání člověka s počátkem, jenž je posvátný, proto je zde ve hře jiná úloha času než je čas pochopený jako fyzikální entita. Současnost potřebuje vytržení tohoto typu, protože je nasměrována jen směrem k budoucnosti a takové rozvržení života dává přednost jen práci bez přestání, takový život je někdy velmi těžký a to přispívá ke vzniku jevů, kterým říkáme 'útěky ze společnosti'. Tyto útěky se realizují formou drog, alkoholu, práce a prevalence osobní výkonnosti. [The article concentrates on the phenomenon of festivity, which is inherent to every championship. The festivity is analyzed as phenomenon of human meeting with an origin, that is sacred and therefore we find here the other role of time as in the physical science. Present needs an ecstasy from this point of view, because is directed only to the future and this scheme prefers to persistent working, such life is very difficult and supports to the inception of phenomena, that are named the escapes from society. This escapes are realised through drugs, alcohol, workoholism anf through limitless personal performance.

  8. Cvičení jako pozitivní coping žen v kontextu změn rodinného života [Physical activity as a source of the positive coping of women in the context of present family life changes

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    Jana Harvanová

    2008-03-01

    Full Text Available Práce zkoumá na vzorku 48 žen ve věku střední a pozdní dospělosti, jak se mateřství a rodinný stav promítá do jednotlivých oblastí životní spokojenosti. Potvrzuje se význam demografi ckých faktorů věku, rodinného stavu a fertility jako možných stresorů žen v kontextu změn současné rodiny. Z dílčích výsledků Dotazníku životní spokojenosti (DŽS a Strategie zvládání stresu (SVF se prokazuje, že stresory mající vliv na prožívání žen, mohou být zmírňovány pravidelnou pohybovou aktivitou (PPA. Cvičení a prokázaná adherence k pravidelné pohybové aktivitě jsou zdrojem využitelným pro pozitivní coping umožňující adaptaci. Ve zkoumaném vzorku se prokázaly tendence mezi mírou adherence a pozitivní zvládací strategií typu Pozitivní sebeinstrukce a Podhodnocení. Tuto zvládací strategii preferují ženy s vyšší mírou adherence ke cvičení. [The study examines how motherhood and family situation refl ex itself in several sections of life satisfaction in a sample of 48 women in middle and late adulthood. The roles of demographical factors of age, family situation and fertility as stressors in women in the context of changes in family life have been confi rmed. The data obtained from the Life Satisfaction Questionnaire and the Stress Coping Strategy questionnaire show that stressors infl uencing women's experiencing can be reduced by regular physical activity. Exercising and proved adherence to regular physical activity can serve as a source for positive coping and allow adaptation. In the sample, an association between the level of adherence to physical activity and positive coping strategies such as Positive self-instruction and Underestimation has been proved. Women showing higher level of adherence to physical activity prefer these types of coping strategies.

  9. Atopowe zapalenie skóry – ciężka postać u 7-letniej dziewczynki – opis przypadku

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    Bolesław Kalicki

    2011-12-01

    Full Text Available Atopowe zapalenie skóry (AZS jest przewlekłym schorzeniem dermatologicznym o podłożu zapalnym, charakteryzującym się częstymi nawrotami zmian, wybitną suchością skóry oraz towarzyszącym świądem. Zapadalność na AZS w ostatnich latach rośnie, aktualnie dotyka ono 1% do 20% populacji dziecięcej. W patogenezie choroby rolę odrywają czynniki genetyczne, środowiskowe, infekcyjne oraz mechanizmy immunologiczne. W zależności od wieku dziecka wyróżniamy fazę niemowlęcą z typowym zajęciem skóry twarzy, tułowia i wyprostnych części kończyn, fazę dziecięcą z przewagą zmian w zgięciach łokciowych, podkolanowych, wokół nadgarstków i kostek oraz przewlekłą postać typu dorosłego. Początkowo zmiany mają charakter sączących się grudek na podłożu rumieniowym, wraz z długością trwania choroby tworzą się blaszki, a naskórek ulega lichenizacji. Podstawą do rozpoznania AZS jest obraz kliniczny. Najczęściej stosowane są w tym celu kryteria Hanifina i Rajki omówione poniżej, a także kryteria zaproponowane w 2003 roku przez American Academy of Dermatology. Chorobie towarzyszy często eozynofilia, zwiększone stężenie immunoglobulin E oraz dodatnie punktowe testy skórne. W pracy przedstawiono najczęstsze jednostki chorobowe wymagające różnicowania z AZS. Leczenie tego schorzenia obejmuje systematyczne nawilżanie skóry, a w fazie zaostrzenia stosowanie leków przeciwzapalnych – glikokortykosteroidów i inhibitorów kalcyneuryny oraz leków przeciwbakteryjnych i antyhistaminowych. W pracy omówiono również ciężką postać AZS przebiegającą z wybitną eozynofilią i hipergammaglobulinemią E u 7-letniej dziewczynki hospitalizowanej w Klinice Pediatrii WIM.

  10. Topiramat – przegląd wybranych artykułów

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    Andrzej Klimek

    2013-06-01

    Full Text Available Padaczka stanowi poważny problem epidemiologiczny z powodu znacznego rozpowszechnienia, a także terapeutyczny ze względu na odsetek przypadków lekooporności wynoszący 30%. W związku z tym położono nacisk na proces zsyntetyzowania nowych leków przeciwpadaczkowych (LPP, które będzie cechować większa skuteczność niż dotychczasowych. Pomimo wprowadzenia do obrotu ponad dziesięciu LPP drugiej generacji odsetek przypadków padaczki lekoopornej nie uległ zmianie, aczkolwiek u pojedynczych chorych wykazały one wyższość nad preparatami stosowanymi tradycyjnie. Do leków przeciwpadaczkowych drugiej generacji należy między innymi topiramat (TPM, cechujący się unikalną budową chemiczną (pochodna fruktopiranozy. W Polsce został wprowadzony do obrotu w 1998 roku. W niniejszej pracy opisano właściwości farmakokinetyczne, jak również zastosowanie TPM jako terapii dodanej w napadach padaczkowych uogólnionych i napadach częściowych oraz jako monoterapii w tego typu napadach. Topiramat ma wielokierunkowy mechanizm działania. Jest silnym blokerem kanałów sodowych, a także wpływa na kanały wapniowe, zwiększa aktywność receptorów GABA, blokuje typ kainowy/AMPA receptorów glutaminowych, jest słabym inhibitorem anhydrozy węglanowej. Omówiono dodatkowo możliwości zastosowania TPM w status epilepticus. Zwrócono uwagę na koszt leczenia topiramatem w porównaniu z takimi lekami, jak kwas walproinowy, oraz wpływ na jakość życia w porównaniu na przykład z lamotryginą. Opisano objawy uboczne, przyglądając się bliżej przyczynie spadku wagi przy zastosowaniu TPM oraz możliwości wystąpienia kwasicy metabolicznej, kamicy nerkowej oraz zmniejszonej potliwości. Podkreślono zalety stosowania topiramatu na tle innych LPP.

  11. Wpływ leków przeciwpadaczkowych na wskaźniki występowania osteoporozy

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    Agnieszka Karp-Majewska

    2010-06-01

    Full Text Available Osteoporoza jest uogólnioną chorobą szkieletu, charakteryzującą się niską masą kostną i zaburzeniem jego mikroarchitektury, prowadzącą do obniżenia wytrzymałości mechanicznej kości oraz zwiększenia podatności na złamania. Do istotnych czynników prowadzących do osteoporozy należy przyjmowanie leków przeciwpadaczkowych. Padaczka często ujawnia się u osób młodych i wymaga długotrwałego leczenia. Konieczne jest zatem wczesne rozpoznanie zaburzeń w metabolizmie ko- ści i wdrożenie odpowiedniej profilaktyki lub leczenia. Za standard diagnostyki osteoporozy uchodzi obecnie badanie gęstości mineralnej kości (BMD metodą podwójnej absorpcjometrii rentgenowskiej (DEXA. Jest to jednak badanie ujawniające obraz osteoporozy na tyle zaawansowanej, że jej przebieg trudno już odwrócić. Dlatego konieczne jest poszukiwanie nowych, bardziej czułych metod diagnostycznych, takich jak markery obrotu kostnego. Badania zostały przeprowadzone u 100 pacjentów w wieku 20-50 lat chorujących na padaczkę powyżej 2 lat oraz 40 osób z grupy kontrolnej dobranych odpowiednio pod wzglę- dem wieku i płci, nieobciążonych innymi schorzeniami. Pacjentom tym wykonywano DEXA oraz oznaczenia markerów obrotu kostnego w surowicy krwi: C-końcowego usieciowanego telopeptydu łańcucha alfa kolagenu typu I (CTX oraz osteokalcyny. Wnioskuje się, iż stosowanie przewlekle leków przeciwpadaczkowych w znaczący sposób wpływa na obniżenie gęstości tkanki kostnej. Zależność ta jest proporcjonalna do długości stosowania leków oraz ich rodzaju. Najbardziej czułym markerem w rozpoznawaniu zaburzeń metabolizmu kostnego u pacjentów leczonych AED poniżej 6 lat jest białko CTX, zaś u osób leczonych dłużej niż 6 lat – DEXA. Największe zmiany tkanki kostnej stwierdzono u pacjentów leczonych fenytoiną.

  12. Lewetiracetam i leczenie padaczki w szczególnych sytuacjach klinicznych

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    Karol Jastrzębski

    2013-12-01

    Full Text Available Padaczka (epilepsia to przewlekłe i częste schorzenie neurologiczne, wymagające długotrwałego stosowania leków przeciwpadaczkowych. Liczne i złożone objawy tego skomplikowanego procesu patofizjologicznego są wynikiem różnych zaburzeń funkcji mózgu. W leczeniu padaczki nie występuje jeden standardowy sposób postępowania. Głównym i podstawowym celem leczenia padaczki jest całkowita kontrola napadów i zminimalizowanie objawów niepożądanych spowodowanych terapią lekami przeciwpadaczkowymi. Ogromne znaczenie ma indywidualne dopasowanie leku do każdego pacjenta. Lek powinien być dostosowany do typu napadu lub zespołu padaczkowego, czę- stości i ciężkości napadów. Pojawienie się leków nowej generacji dało im pewną przewagę w stosunku do leków starszej generacji. Jednym z dostępnych na rynku leków przeciwpadaczkowych jest lewetiracetam, występujący w postaci podłużnych tabletek. Można go stosować w monoterapii u pacjentów od 16. roku życia z nowo rozpoznaną padaczką, w leczeniu napadów padaczkowych o początku ogniskowym z wtórnym uogólnieniem lub bez wtórnego uogólnienia. Lewetiracetam można również dodawać do terapii innymi lekami przeciwpadaczkowymi, w leczeniu m.in. napadów padaczkowych o początku ogniskowym z wtórnym uogólnieniem lub bez wtórnego uogólnienia u dzieci od 1 miesiąca życia, drgawek mioklonicznych (u pacjentów w wieku powyżej 12 lat z młodzieńczą padaczką miokloniczną czy pierwotnych uogólnionych napadów toniczno-klonicznych u pacjentów powyżej 12 lat z idiopatyczną padaczką uogólnioną. Lek ten między innymi pomaga w stabilizowaniu aktywności elektrycznej w mózgu i zapobiega napadom.

  13. Rola białka AS160/TBC1D4 w transporcie glukozy do wnętrza miocytów[i][/i

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    Agnieszka Mikłosz

    2011-01-01

    Full Text Available Mięśnie szkieletowe to jedne z najważniejszych tkanek uczestniczących w utrzymaniu homeostazy glukozy całego organizmu. Glukoza przenika do komórek mięśniowych na zasadzie dyfuzji ułatwionej, zachodzącej z udziałem transporterów glukozy (GLUT. Stymulacja insulinowego (związanego z aktywacją 3-kinazy fosfatydyloinozytolu – PI3K bądź też niezależnego od insuliny (związanego z aktywacją kinazy zależnej od AMP – AMPK szlaku przekaźnictwa sygnału uruchamia kaskadę reakcji prowadzącą do translokacji GLUT-4 do błony komórkowej, a w konsekwencji do wzrostu wychwytu glukozy w miocytach. W prowadzonych ostatnio badaniach wykazano, że bezpośrednio w proces translokacji GLUT-4 jest zaangażowane białko sygnałowe określane jako AS160 – substrat Akt o masie cząsteczkowej 160 kDa. Białko to prawdopodobnie jest ogniwem łączącym szlak insulinowy ze szlakiem zależnym od aktywacji kinazy AMPK. Badania potwierdzają, iż fosforylacja AS160 ulega wzmożeniu zarówno pod wpływem stymulacji insuliną, jak też podczas wysiłku fizycznego, co wskazuje na ich addytywną zależność. W mięśniach szkieletowych osób z opornością na insulinę i/lub cukrzycą typu 2 dochodzi do znacznego obniżenia zależnej od insuliny fosforylacji białka AS160 i spadku translokacji do błony GLUT-4. Stąd też zmniejszony poziom insulinozależnej fosforylacji AS160 może odgrywać istotną rolę w oporności na insulinę[i] in vivo[/i].

  14. Depresja u chorych na reumatoidalne zapalenie stawów

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    Brygida Kwiatkowska

    2011-04-01

    Full Text Available Depresja jest najczęstszym schorzeniem psychicznym występującymu pacjentów tzw. podstawowej opieki zdrowotnej i stwierdzasię ją u 12,5% chorych, a w przebiegu schorzeń przewlekłych jestroz poznawana wielokrotnie częściej.Do rozwoju depresji w przebiegu schorzeń przewlekłych przyczyniasię wiele czynników, takich jak: predyspozycje genetyczne, predyspozycjepsychiczne, długotrwale utrzymujący się stres orazzwiększenie stężeń cytokin prozapalnych w ośrodkowym układzienerwowym. Ośrodkowy układ nerwowy w wyniku działania cytokinprozapalnych aktywuje wiele zmian w układzie neuroendokrynnymi immunologicznym, określanych jako „zachowanie chorobowe” (sickness behaviour.W reumatoidalnym zapaleniu stawów (RZS depresja występujeu 13–65% chorych, ale tylko u 25% chorych jest rozpoznana. Tak częstewystępowanie depresji jest wynikiem utrzymywania się dużegostężenia cytokin prozapalnych wtym schorzeniu, powodującego zaburzeniaw zachowaniu. Depresja wpływa negatywnie na przebieg RZS,nasilając objawy somatyczne choroby. Z kolei wyniki leczenia depresjisą tym gorsze, im większa jest aktywność RZS, co powoduje zmniejszenieskuteczności leczenia podtrzymującego i zwiększa możliwośćwy stąpienia nawrotów depresji po pierwszym epizodzie. Udowadniato konieczność jednoczesnego intensywnego leczenia obu chorób i ichwczesnego rozpoznania. Rozpoznanie depresji może nastręczać dużotrudności z powodu nakładania się objawów wy stępujących zarównow depresji, jak i w RZS, takich jak: zmęczenie, brak apetytu, zmniejszeniemasy ciała, różnego typu dolegliwości bólowe itp. Nierozpoznanadepresja w przebiegu RZS, poza wpływem na nasilenie objawówchoroby, zmniejsza skuteczność stosowanej terapii, z uwagi na gorsząwspółpracę pacjenta i mniejsze zaangażowanie w leczenie. Zwiększato ryzyko myśli i prób samobójczych oraz skraca czas przeżycia pacjentówchorych na RZS.

  15. Trening uwagi u dzieci z ADHD – przegląd badań

    Directory of Open Access Journals (Sweden)

    Monika Deja

    2017-03-01

    Full Text Available Jednym z zagadnień, które w ostatnich latach wzbudziły zainteresowanie psychologów poznawczych i rozwojowych, jest trening poznawczy oraz związane z nim zjawisko transferu pozytywnego. Okazuje się bowiem, że nawet krótkotrwały trening poznawczy prowadzi zarówno do poprawy funkcjonowania ćwiczonego procesu, jak i do transferu efektów na funkcje, które nie były ćwiczone. Badacze z jednej strony poszukują metod treningu dających najlepsze i najtrwalsze rezultaty, a z drugiej zastanawiają się nad implikacjami praktycznymi tego typu oddziaływań. Grupą chętnie badaną i poddawaną interwencjom poznawczym są dzieci z zespołem nadpobudliwości psychoruchowej z deficytem uwagi. W niniejszej pracy przedstawiono przegląd badań nad treningami uwagi u dzieci z tej grupy klinicznej. Celem przeglądu było przeanalizowanie skuteczności różnych treningów uwagi i zakresu transferu efektów na inne, nietrenowane funkcje poznawcze. Zaprezentowane doniesienia dowodzą, iż treningi uwagi zmniejszają intensywność objawów nieuwagi, impulsywności i nadaktywności u dzieci z zespołem nadpobudliwości psychoruchowej z deficytem uwagi. Potwierdzają  to  zarówno wyniki zastosowanych testów diagnostycznych, jak i badania EEG czy opinie rodziców. Ponadto zaobserwowano poprawę sprawności różnych funkcji uwagi oraz innych funkcji poznawczych (w tym pamięci operacyjnej, a nawet inteligencji płynnej. Wyniki napawają optymizmem, gdyż dowodzą, że możliwe jest zmniejszenie nasilenia objawów ADHD u dzieci w stosunkowo krótkim czasie. Niewątpliwie są też zachętą do  dalszych badań w tym zakresie.

  16. Pęknięcie nerki jako rzadkie powikłanie zabiegu ESWL (kruszenie kamieni falą generowaną pozaustrojowo

    Directory of Open Access Journals (Sweden)

    Tomasz Ząbkowski

    2010-12-01

    Full Text Available Kruszenie kamieni falą generowaną pozaustrojowo (extracorporeal shock wave lithotripsy, ESWL należy do podstawowych metod leczenia kamicy nerkowej – zyskała ona szerokie zastosowanie jako metoda małoinwazyjna i skuteczna. Wraz ze wzrostem doświadczenia urologów wykonujących zabiegi litotrypsji oraz udoskonalaniem litotryptorów wzrasta skuteczność wykonywanych zabiegów. Autorzy przedstawiają niezmiernie rzadkie powikłanie po zabiegu ESWL pod postacią pęknięcia nerki i masywnego krwawienia do przestrzeni zaotrzewnowej. Siedemdziesięcioośmioletni pacjent z małopłytkowością, po zabiegu by-passów (coronary artery bypass graft, CABG, po implantacji układu stymulującego serca, z nadciśnieniem tętniczym i cukrzycą typu II, leczony doustnymi antykoagulantami został przyjęty do Kliniki Urologicznej 6 godzin po zabiegu ESWL w trybie pilnym z powodu pogarszającego się stanu ogólnego, narastających dolegliwości bólowych w lewej okolicy lędźwiowej i masywnego krwiomoczu. Po przeprowadzeniu badania podmiotowego, przedmiotowego i na podstawie wykonanych badań dodatkowych (tomografia komputerowa jamy brzusznej rozpoznano pęknięcie nerki z masywnym krwiakiem przestrzeni zaotrzewnowej. Chorego zakwalifikowano w trybie pilnym do leczenia operacyjnego. Śródoperacyjnie stwierdzono ogromny krwiak wypełniający przestrzeń zaotrzewnową wokół nerki lewej i schodzący do lewego dołu biodrowego oraz masywne krwawienie z całej powierzchni uszkodzonej nerki bez możliwości opanowania krwawienia. Wobec powyższego obrazu podjęto decyzję o wycięciu nerki lewej. Z uwagi na stwierdzone płaszczyznowe krwawienie w otaczających tkankach i brak pewności co do ostatecznej hemostazy zdecydowano wykonać packing – założono 6 serwet do przestrzeni zaotrzewnowej. We wczesnym okresie pooperacyjnym nie stwierdzono powikłań. Pacjent krążeniowo, oddechowo i nerkowo wydolny. W 3. dobie od zabiegu w trybie planowym usuni

  17. Synthesis, Microstructure and the Crystalline Structure of the Barium Titanate Ceramics Doped with Lanthanum

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    Wodecka-Duś B.

    2013-12-01

    Full Text Available W prezentowanej pracy przeprowadzono badania ceramiki BaTiO3 i Ba1-xLąxTi1-x/4O3 (BLT dla koncentracji z prze- działu 0,001< x <0,004 (0,l-0,4mol.% La. Ceramikę BLT wytworzono z mieszaniny prostych tlenków La203, TiOi i BaCOj (wszystkie o czystości 99,9+%, Aldrich Chemical Co. Proszki ceramiczne otrzymano metodą konwencjonalną w stanie stałym (metodą MOM i poddano badaniu mikrostruktury i struktury krystalicznej. Mieszaniny proszków poddano analizie termicznej. Wyniki analizy termicznej określiły optymalną temperaturę syntezy oraz procesy zachodzące podczas ogrzewania proszków. Następnie proszki formowano w dyski pod ciśnieniem 300MPa w matrycach ze stali nierdzewnej o średnicy 10 mm. Syntezę przeprowadzono w Ts =950°C t =2godz. Ostatnim krokiem technologii było bezciśnieniowe spiekanie metodą swobodnego spiekania w T = 1350^ przez / =2 godziny. Morfologię otrzymanego materiału ceramicznego obserwowano metodą skaningowej mikroskopii elektronowej. Ceramikę BLT badano również pod względem składu chemicznego metodą EDS. Analizę strukturalną przeprowadzono metodą dyfrakcji rentgenowskiej. Badania mikrostruktury i struktury krystalicznej ceramiki przeprowadzono w temperaturze pokojowej. Badania EDS potwierdziły zachowanie stechiometrii otrzymanych próbek według wzoru chemicznego. Rentgenowska analiza dyfrakcyjna potwierdziły wytworzenie pożądanej struktury krystalicznej zarówno czystej ceramiki BaTiOj jak i z domieszką Lau. Otrzymana ceramika wykazuje strukturę typu perowskitu A BO? o symetrii tetragonalnej P4 mm. Stwierdzono, że wraz ze wzrostem stężenia La3* w BaTiOj następuje zmniejszenie wielkości ziam krystalicznych, zmniejszenie średniego wymiaru krystalitów, zmniejszenie objętości komórki elementarnej oraz wzrost obliczonej rentgenowskiej gęstości.

  18. Towards Application of Speech Act Theory to Opinion Mining

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    Agnieszka Magdalena Pluwak

    2016-12-01

    Full Text Available Towards the Application of Speech Act Theory to Opinion Mining The paper refers to the pragmatics’ perspective on opinion mining in Polish and English, inspired by the discrepancy between the coverage of sentiment analysis and the market demand. An analysis of speech acts expressed in opinion texts reveals that almost half of all opinions include ways of indirect evaluation that might not get extracted while applying traditional methods of sentiment analysis based on direct evaluative vocabulary and polarity lexicons. Coding of sentiment with respect to speech acts could vastly broaden data mining results within an NLP-system.   O zastosowaniu teorii aktów mowy w ekstrakcji danych z tekstów opinii internetowych Jedno z aktualnych zagadnień językoznawstwa komputerowego, jakim jest automatyczne badanie wydźwięku wypowiedzi, nie uwzględniło dotychczas w wystarczającym stopniu pragmatyki językoznawczej, np. aktów mowy Austina (1961 i Searla (1969, a zatem również implicytnych sposobów wyrażania ewaluacji. Tymczasem podejście od pragmatyki ku konstrukcjom przełożonym na reguły programistyczne umożliwiłoby nie tylko szersze spojrzenie na analizę sentymentu, ale też zbliżyłoby automat do sposobu, w jaki odbiera go człowiek. W szczególności chodzi tu sposoby wyrażania (niezadowolenia wykraczające poza poziom leksykalny (bez nacechowanej negatywnie leksyki, typu Nigdy więcej tam nie pójdę. Artykuł prezentuje: 1. aktualne podejścia do analizy wydźwięku w lingwistyce komputerowej, 2. propozycję zastosowania podejścia pragmatycznego, 3. wyniki badania próbki tekstów opinii internetowych pod kątem występowania w nich aktów mowy, 4. propozycję utworzenia reguł ekstrakcji danych na ich podstawie. Zaprezentowane podejście zakłada hipotezę wtórnej oralności, czyli tego, że język opinii jest zapisanym językiem mówionym.

  19. Klindamycyna – kompletna monografia leku

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    Iwona Korzeniewska-Rybicka

    2018-01-01

    Full Text Available Klindamycyna jest półsyntetyczną chlorową pochodną linkomycyny o licznych zaletach w porównaniu ze swoim prekursorem, dostępną w lecznictwie od 1966 roku. Wykazuje działanie bakteriostatyczne poprzez wpływ na podjednostkę 50S rybosomów wrażliwych bakterii, identycznie jak makrolidy i streptograminy typu B, a co nie mniej ważne, blokuje syntezę toksyn i ogranicza wirulencję niezwykle niebezpiecznych bakterii: Streptococcus pyogenes i Staphylococcus aureus. Jest to antybiotyk aktywny wobec ziarenkowców Gram-dodatnich, z wyjątkiem enterokoków, oraz wobec większości bakterii beztlenowych, zarówno Gram-dodatnich, jak i Gram-ujemnych. Działa też przeciwpierwotniakowo, np. wobec gatunków takich jak Pneumocystis czy Toxoplasma. Oporność bakterii na klindamycynę narasta wśród niektórych gatunków, np. Streptococcus pneumoniae czy Bacteroides fragilis, co związane jest zwykle z modyfikacją miejsca docelowego wiązania, działaniem pomp wyrzutu lub inaktywacją enzymatyczną leku. Klindamycyna jest dostępna w Polsce pod różnymi postaciami. Charakteryzuje się dobrą biodostępnością po podaniu doustnym oraz dobrą dystrybucją do tkanek (z wyjątkiem ośrodkowego układu nerwowego. Eliminowana jest głównie z żółcią. Działania niepożądane klindamycyny na ogół nie są groźne, choć stosunkowo często występują objawy takie jak biegunka czy wysypki skórne. Poważnym powikłaniem terapii tym antybiotykiem może być wystąpienie rzekomobłoniastego zapalenia jelit wywołanego przez Clostridium difficile. Klindamycyna może być stosowana w wielu wskazaniach klinicznych. Często jest jednak lekiem drugiego rzutu w terapii zakażeń lub w farmakoprofilaktyce. Aktywność wobec bakterii beztlenowych i gronkowców opornych na metycylinę przesuwa ją do terapii pierwszej linii leczenia w infekcjach o takiej etiologii.

  20. Wyrazy z interfiksem -ø- (formantem zerowym w języku serbskim (analiza słowotwórczo-semantyczna

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    Dragana M. Ratković

    2017-12-01

    Full Text Available Words with the interfix -ø- (the zero interfix in the Serbian language The work represents a derivational and semantic analysis of the words with the zero interfix in the contemporary Serbian language in accordance with the conceptual and terminological apparatus of modern Slavic derivatology. The author argues that the zero interfix, as the differentia specifica of compound words, occurred in the Serbian language very early – at the end of the 12th century and is currently very productive because of how economical the lexemes thus created are. In modern Serbian, new words of this type appear by way of borrowing from other languages, primarily from the English language (e.g., pank-moda, boks-meč, pres-centar etc..   Wyrazy z interfiksem -ø- (formant zerowy w języku serbskim (analiza słowotwórczo-semantyczna Praca stanowi słowotwórczo-semantyczną analizę wyrazów z interfiksem zerowym we współczesnym języku serbskim zgodnie z pojęciowoterminologicznym aparatem współczesnego słowotwórstwa slawistycznego. Autorka wskazuje również na to, że zerowy interfiks, jako differentia specifica wyrazów powstałych przez złożenie, w języku serbskim pojawia się bardzo wcześnie – istnieje już w końcu XII wieku. Aktualnie model słowotwórczy wyrazów z interfiksem zerowym jest niezwykle produktywny ze względu na ekonomiczność leksemów tego typu. We współczesnym języku pojawiają się one przede wszystkim za sprawą zapożyczeń z innych języków, głównie z angielskiego (np. pank-moda, boks-meč, pres-centar itd..

  1. Modernistyczny terror, czyli krótko o "tunnel vision" Jamesa C. Scotta

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    Mateusz Pietryka

    2014-06-01

    Full Text Available The terror of modernism or Scott’s 'tunnel vision'  The article analyses James C. Scott’s book Seeing Like a State: How Certain Schemes to Improve the Human Condition Have Failed. Scott’s book examines different aspects of a modern state’s activity and its methods of control over the population. Scott describes diverse failures in state planning and links them to the ideology of what he terms “high modernism”. Scott states that this ideology results in over confidence in scientific progress and omnipresent simplifications, since a modernist plan does not recognize the importance of local knowledge and tradition. Though the book clearly illustrates how the fore-mentioned centralist approach leads to failures, its methodology, however, lacks the critical analysis of the nature of governance and liberal economy. Furthermore, Scott offers no insight on the late capitalism nor legible solutions for the described issues. He seems to be unaware of constructing his own “tunnel vision” by selective case studies and building the narration on simplified oppositions.   Modernistyczny terror, czyli krótko o tunnel vision Jamesa C. Scotta Artykuł jest analizą książki Jamesa Scotta, Seeing Like a State: How Certain Schemes to Improve the Human Condition Have Failed. Książka ta bada różnorodne aspekty aktywności współczesnego państwa i jego metody kontroli nad populacją. Autor opisuje rożnego typu niepowodzenia wynikające z państwowego planowania i wiąże je z ideologią nazwaną przez siebie „zaawansowanym modernizmem”. Scott twierdzi, że skutkuje ona pokładaniem nadmiernej wiary w postęp naukowy i wszechobecnymi uproszczeniami, ponieważ modernistyczny plan nie uznaje znaczenia lokalnej wiedzy i tradycji. Choć książka jasno pokazuje, jak wspomniane centralistyczne podejście kończy się porażkami, jej metodologii brakuje krytycznej analizy natury władzy i liberalnej ekonomii. Ponadto Scott pomija temat p

  2. Zastosowanie programów komputerowych w rehabilitacji neuropsychologicznej dysfunkcji poznawczych u pacjentów ze stwardnieniem rozsianym

    Directory of Open Access Journals (Sweden)

    Ernest Tyburski

    2013-06-01

    Full Text Available W obrazie klinicznym stwardnienia rozsianego (łac. sclerosis multiplex, SM na funkcjonowanie chorych wpływają – poza objawami neurologicznymi – współwystępujące objawy neuropsychologiczne, do których zalicza się zaburzenia emocjonalne i dysfunkcje poznawcze oraz czynniki osobowościowe. Dysfunkcje w sferze procesów uwagi, funkcji wykonawczych czy pamięci mają wpływ na zmniejszenie zdolności adaptacyjnych, które są kluczowe dla jakości życia chorych. W badaniach dużych grup klinicznych udowodniono obecność dysfunkcji poznawczych u 40–65% pacjentów. Najczęściej na SM zapadają osoby młode, u których dysfunkcje ruchowe i zaburzenia poznawcze mogą utrudniać codzienne funkcjonowanie, a często też stają się przeszkodą w podejmowaniu zadań życiowych. Dlatego też istnieje potrzeba opracowania nowych i skutecznych programów rehabilitacyjnych dla tej grupy chorych. Rehabilitacja neuropsychologiczna pacjentów z SM obejmuje różnego rodzaju oddziaływania, których celem jest leczenie dysfunkcji poznawczych. W pracy neuropsychologa coraz częściej jako narzędzie terapeutyczne wykorzystuje się programy komputerowe służące do treningów poznawczych. Podstawą dla tego typu oddziaływań są dowody świadczące o zmianach neuroplastycznych u osób ze stwardnieniem rozsianym. Największe efekty terapeutyczne osiąga się jednak dzięki współpracy zespołu interdyscyplinarnego, w którego skład powinni wchodzić neurolog, psychiatra, neuropsycholog oraz rehabilitant. W Polsce uzyskanie takiej pomocy przez pacjentów z SM jest nadal bardzo trudne. Przykładem obrazującym skuteczne zastosowanie rehabilitacji neuropsychologicznej za pomocą programów komputerowych jest studium przypadku chorego ze stwardnieniem rozsianym.

  3. Influence of operating and water-chemistry parameters on fuel cladding corrosion and deposition of corrosion products on cladding surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Kritsky, V.G.; Berezina, I.G.; Rodionov, Y.A., E-mail: kritsky@givnipiet.spb.ru, E-mail: alemaskina@givnipiet.ru [Leading Inst. ' VNIPIET' , Saint Petersburg (Russian Federation)

    2010-07-01

    A description of mass transfer mechanism of corrosion products in the primary coolant circuit is a complicated problem. The deposits of crud is to be proportional to the amount of corrosion products circulating in the primary coolant circuit, therefore all models of mass transfer in the circuit include the change of corrosion products concentration and the corrosion rate in time, removing these products by filters and deposition. Decontamination of the circuit equipment and replacement work needs lead to a local change of corrosion rate which results in the increase of corrosion products concentration in the circuit and the increase of deposits on surfaces. If due to incorrect water chemistry conditions for corrosion products deposition in the core are created not only the activity of the coolant increases but the hydraulic resistance of the reactor also grows which results in the increase of the pressure drop at the reactor. The phenomenon of 'pressure drop' which takes place in NPP with VVER reactors was considered. The reasons of this phenomenon are the following: the great removal of corrosion products (CP) from steam generator surfaces after decontamination, change of CP behavior and then consequent deposit of CP on the fuel element surfaces; and, sub-cooled boiling takes place on the some of fuel element and results in the acceleration of corrosion products deposit, the increase of nuclide activation period and coolant radioactivity. A model was developed to explain pressure drop rise in-core and deposits redistribution in the core and in the primary circuit of NPP with VVER-440. The physical-chemical basis of the model is the transport corrosion products dependence on temperature, pH{sub T} value of coolant, and correlation between rates of corrosion products (Fe) formation (after steam generators decontamination) and their removing from the circuit. The aim of our modeling is to predict the growth of pressure difference on the basis of regular

  4. VISWAM. A computer code package for thermal reactor physics computations

    Energy Technology Data Exchange (ETDEWEB)

    Jagannathan, V.; Thiyagarajan, T.K.; Ganesan, S.; Jain, R.P.; Pal, U. [Bhabha Atomic Research Centre, Mumbai (India); Karthikeyan, R. [Ecole Polytechnique de Montreal, Montreal, Quebec (Canada)

    2004-07-01

    The nuclear cross section data and reactor physics design methods developed over the past three decades have attained a high degree of reliability for thermal power reactor design and analysis. This is borne out from the analysis of physics commissioning experiments and several reactor-years of operational experience of two types of Indian thermal power reactors, viz. BWRs and PHWRs. Our computational tools were also developed and tested against a large number of IAEA CRP benchmarks on in-core fuel management code package validation for the modern BWR, PWR, VVER and PHWR. Though the computational algorithms are well tested, their mode of use has remained rather obsolete since the codes were developed when the modern high-speed large memory computers were not available. The use of Fortran language limits their potential use for varied applications. We are developing specific Visual Interface Software as the Work Aid support for effective Man-Machine interface (VISWAM). The VISWAM package when fully developed and tested will enable handling the input description of complex fuel assembly and the reactor core geometry with immaculate ease. Selective display of the three dimensional distribution of multi-group fluxes, power distribution and hot spots will provide a good insight into the analysis and also enable inter comparison of different nuclear datasets and methods. Since the new package will be user-friendly, training of requisite human resource for the expanding Indian nuclear power programme will be rendered easier and the gap between an expert and a novice will be greatly reduced. (author)

  5. Problems and prospects of nuclear power plants construction

    Directory of Open Access Journals (Sweden)

    Pergamenshhik Boris Klimentyevich

    2014-02-01

    Full Text Available 60 years ago, in July 1954 in the city of Obninsk near Moscow the world's first nuclear power plant was commissioned with a capacity of 5 MW. Today more than 430 nuclear units with a total capacity of almost 375000 MW are in operation in dozens of the countries worldwide. 72 electrical power units are currently under construction, 8 of them are located in the Russian Federation. There will be no alternative to nuclear energy in the coming decades. Among the factors contributing to the construction of nuclear power plants reckon limited fossil fuel supply, lack of air and primarily carbon dioxide emissions. The holding back factors are breakdown, hazard, radioactive wastes, high construction costs and long construction period. Nuclear accidents in the power plant of «Three-Mile-Island» in the USA, in Chernobyl and in Japan have resulted in termination of construction projects and closure of several nuclear power plants in the Western Europe. The safety systems have become more complex, material consumption and construction costs have significantly increased. The success of modern competing projects like EPR-1600, AP1000, ABWR, national ones AES-2006 and VVER-TOI, as well as several others, depends not only on structural and configuration but also on construction and technological solutions. The increase of the construction term by one year leads to growth of estimated total costs by 3—10 %. The main improvement potentials include external plate reinforcement, pre-fabricated large-block assembly, production and installation of the equipment packages and other. One of the crucial success factors is highly skilled civil engineers training.

  6. Development of an Analytic Nodal Diffusion Solver in Multi-groups for 3D Reactor Cores with Rectangular or Hexagonal Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, Juan Andres; Aragones, Jose Maria; Garcia-Herranz, Nuria [Universidad Politecnica de Madrid, 28006 Jose Gutierrez Abascal 2, Madrid (Spain)

    2008-07-01

    More accurate modelling of physical phenomena involved in present and future nuclear reactors requires a multi-scale and multi-physics approach. This challenge can be accomplished by the coupling of best-estimate core-physics, thermal-hydraulics and multi-physics solvers. In order to make viable that coupling, the current trends in reactor simulations are along the development of a new generation of tools based on user-friendly, modular, easily linkable, faster and more accurate codes to be integrated in common platforms. These premises are in the origin of the NURESIM Integrated Project within the 6. European Framework Program, which is envisaged to provide the initial step towards a Common European Standard Software Platform for nuclear reactors simulations. In the frame of this project and to reach the above-mentioned goals, a 3-D multigroup nodal solver for neutron diffusion calculations called ANDES (Analytic Nodal Diffusion Equation Solver) has been developed and tested in-depth in this Thesis. ANDES solves the steady-state and time-dependent neutron diffusion equation in three-dimensions and any number of energy groups, utilizing the Analytic Coarse-Mesh Finite-Difference (ACMFD) scheme to yield the nodal coupling equations. It can be applied to both Cartesian and triangular-Z geometries, so that simulations of LWR as well as VVER, HTR and fast reactors can be performed. The solver has been implemented in a fully encapsulated way, enabling it as a module to be readily integrated in other codes and platforms. In fact, it can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. Verification of performance has shown that ANDES is a code with high order definition for whole core realistic nodal simulations. In this paper, the methodology developed and involved in ANDES is presented. (authors)

  7. OECD/NEA BENCHMARK FOR UNCERTAINTY ANALYSIS IN MODELING (UAM FOR LWRS – SUMMARY AND DISCUSSION OF NEUTRONICS CASES (PHASE I

    Directory of Open Access Journals (Sweden)

    RYAN N. BRATTON

    2014-06-01

    Full Text Available A Nuclear Energy Agency (NEA, Organization for Economic Co-operation and Development (OECD benchmark for Uncertainty Analysis in Modeling (UAM is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the “Neutronics Phase”, which is devoted mostly to the propagation of nuclear data (cross-section uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: 238U radiative capture and inelastic scattering (n, n’ as well as the average number of neutrons released per fission event of 239Pu.

  8. Latest findings from the OECD Rasplav Project

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.

    1997-01-01

    During the late phase of a severe accident in a light water reactor (current and future designs of BWRs, PWRs and VVERs), a significant amount of core material may relocate downward to the lower head of the reactor vessel. If molten core materials were to relocate to the lower head of the reactor pressure vessel (RPV), a molten pool consisting primarily of a mixture of ZrO{sub 2} and UO{sub 2} and some combination of a metal would form on the lower head. A solid crust of material would form around the boundaries of the pool, but internal heat generation resulting from radioactive decay of fission products would assure that most of the pool remains molten. In fact, the molten pool would undergo significant internal natural convection which would reach steady state conditions in about a few hours. Detailed understanding of all aspects of this natural convection process, in conjunction with the thermal boundary conditions imposed on the outer surface, determines the fraction of the total heat dissipation that is transferred through the upper crust to the inside of the reactor vessel by radiative heat exchange and the fraction which must be conducted through the wall of the reactor vessel lower head. This distribution is critical in determining whether and under what conditions the molten material can be cooled and retained in the reactor pressure vessel. The OECD Rasplav Project was established in 1994 as a three year program to study molten pool behavior and its interactions with structural materials in the lower head. This paper reviews the establishment of the project, its initial studies and proposed experimental testing, and the construction, preparation, and actual testing of a chamber of corium heated to well above liquid temperature.

  9. Adaptive Neural Network Algorithm for Power Control in Nuclear Power Plants

    Science.gov (United States)

    Masri Husam Fayiz, Al

    2017-01-01

    The aim of this paper is to design, test and evaluate a prototype of an adaptive neural network algorithm for the power controlling system of a nuclear power plant. The task of power control in nuclear reactors is one of the fundamental tasks in this field. Therefore, researches are constantly conducted to ameliorate the power reactor control process. Currently, in the Department of Automation in the National Research Nuclear University (NRNU) MEPhI, numerous studies are utilizing various methodologies of artificial intelligence (expert systems, neural networks, fuzzy systems and genetic algorithms) to enhance the performance, safety, efficiency and reliability of nuclear power plants. In particular, a study of an adaptive artificial intelligent power regulator in the control systems of nuclear power reactors is being undertaken to enhance performance and to minimize the output error of the Automatic Power Controller (APC) on the grounds of a multifunctional computer analyzer (simulator) of the Water-Water Energetic Reactor known as Vodo-Vodyanoi Energetichesky Reaktor (VVER) in Russian. In this paper, a block diagram of an adaptive reactor power controller was built on the basis of an intelligent control algorithm. When implementing intelligent neural network principles, it is possible to improve the quality and dynamic of any control system in accordance with the principles of adaptive control. It is common knowledge that an adaptive control system permits adjusting the controller’s parameters according to the transitions in the characteristics of the control object or external disturbances. In this project, it is demonstrated that the propitious options for an automatic power controller in nuclear power plants is a control system constructed on intelligent neural network algorithms.

  10. Steam generator tube failures

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service.

  11. Validation of the fast-running in-vessel model ASTRID for predicting the radioactive releases to the containment

    Energy Technology Data Exchange (ETDEWEB)

    Jaakko, M. [VTT Processes (Finland); Schmuck, P. [Forschungszentrum Karlsruhe, Karlsruhe (Germany)

    2004-07-01

    The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) process model is used for the faster than real-time prediction of the radioactivity released into the containment and further into the environment in case of an emergency situation in a light water reactor. Combined together with the containment module COCOSYS the model can predict the entire radioactivity release chain from the primary system to the containment and further into the environment. In the paper the ASTRID thermohydraulic module PROCESS is presented shortly. The thermohydraulic part is a fast running solution for the drift-flux based thermohydraulics. In high temperatures the core degradation leading to the melt pool formation in the reactor barrel and reactor vessel lower head is calculated in the in-vessel module RELOMEL. Finally after the reactor vessel wall has been eroded due to the molten corium in the lower plenum, the massive radioactivity release occurs into the containment. But even before this scenario the radioactivity may be transported from the superheated core to the containment by the coolant. The reference plants for the development have been the Westinghouse type 4-loop PWR, the French type 3-loop PWR, The German type 4-loop Konvoi PWR, the Loviisa VVER type PWR, and the Olkiluoto type internal pump BWR. The reference code for the DBA thermal hydraulics has been the SMABRE code. In the developmental assessment the capability of the rough nodalization of ASTRID has been tested against the SMABRE nodalization describing the plants with 50 - 500 nodes. For the developmental assessment of the in-vessel severe accident the sample cases are calculated with MELCOR. The more thorough validation is based on the internationally known system codes, RELAP5, MELCOR, CATHARE and ATHLET. In the validation the most problematic area is the radioactivity transport into the containment. This part of the validation is done with the integrated code system. (authors)

  12. Evolution of the CYCLE code for the system analysis of the nuclear fuel cycle

    Directory of Open Access Journals (Sweden)

    A.G. Kalashnikov

    2016-06-01

    Full Text Available The CYCLE code is intended to simulate mathematically the operation of a nuclear power system (NPS with thermal and fast reactors in an open or closed nuclear fuel cycle, to develop scenarios of efficient nuclear power evolution in Russia and to analyze trends in global nuclear power. The code is based on a well-known software program, WIMSD-5B, broadly used for the design of thermal reactor cells, and on a 2D multi-group software system, RZA, for the fast neutron reactor simulation. The CYCLE code was developed at IPPE in Obninsk. This paper presents a brief review of the capabilities and information on the current status of the CYCLE code. The code allows simulation of key facilities of the external fuel cycle (fuel fabrication and reprocessing facilities, SNF storage, uranium, plutonium, neptunium, americium and curium stores, RW long-term storage sites, nuclear reactors, including RBMK-1000 reactors, existing and advanced VVER reactors (using different fuel types, and fast reactors (both existing and innovative. As an important feature, the CYCLE code allows the evolution of the fuel's nuclide composition both in reactors and at the external fuel cycle phase to be considered in details. Offered as an extra option is the capability to calculate a variety of the nuclear fuel cycle cost parameters for nuclear power plants with thermal and fast reactors. For years, the code has been successfully used as part of INPRO, an international innovative nuclear reactor and fuel cycle project. The results of studies into the Russian NPS evolution scenarios were presented at Global 2011. Some other of the CYCLE-based simulation results were presented at Global 2015.

  13. Modified ring stretch tensile testing of Zr-1Nb cladding

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, A.B.; Majumdar, S.; Ruther, W.E.; Billone, M.C.; Chung, H.M.; Neimark, L.A. [Argonne National Lab., IL (United States)

    1998-03-01

    In a round robin effort between the US Nuclear Regulatory Commission, Institut de Protection et de Surete Nucleaire in France, and the Russian Research Centre-Kurchatov Institute, Argonne National Laboratory conducted 16 modified ring stretch tensile tests on unirradiated samples of zr-1Nb cladding, which is used in Russian VVER reactors. Test were conducted at two temperatures (25 and 400 C) and two strain rates (0.001 and 1 s{sup {minus}1}). At 25 C and 0.001 s{sup {minus}1}, the yield strength (YS), ultimate tensile strength (UTS), uniform elongation (UE), and total elongation (TE) were 201 MPa, 331 MPa, 18.2%, and 57.6%, respectively. At 400 C and 0.001 s{sup {minus}1}, the YS, UTS, UE, and TE were 109 MPa, 185 MPa, 15.4%, and 67.7%, respectively. Finally, at 400 C and 1 s{sup {minus}1}, the YS, UTS, UE, and TE were 134 MPa, 189 MPa, 18.9%, and 53.4%, respectively. The high strain rate tests at room temperature were not successful. Test results proved to be very sensitive to the amount of lubrication used on the inserts; because of the large contact area between the inserts and specimen, too little lubrication leads to significantly higher strengths and lower elongations being reported. It is also important to note that only 70 to 80% of the elongation takes place in the gauge section, depending on specimen geometry. The appropriate percentage can be estimated from a simple model or can be calculated from finite-element analysis.

  14. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  15. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  16. Low cycle thermomechanical fatigue of reactor steels: Microstructural and fractographic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Kasl, Josef; Jandova, Dagmar [Výzkumný a zkušební ústav Plzeň s.r.o., Tylova 1581/46, 316 00 Plzen (Czech Republic); Jóni, Bertalan [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Eötvös Loránd University, Egyetem tér 1-3, Budapest H-1053 (Hungary); Misják, Fanni [Centre for Energy Research, Institute of Technical Physics and Materials Science, Konkoly-Thege M. 29-33, Budapest H-1121 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-07-29

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of a VVER-440 reactor pressure vessel were investigated under fully reversed total strain controlled low cycle fatigue tests. The measurements were carried out in isothermal conditions at 260 °C and with thermal-mechanical conditions in the range 150–270 °C using a GLEEBLE-3800 servo-hydraulic thermal-mechanical simulator. The low cycle fatigue results were evaluated with the Coffin–Manson law, and the parameters of the Ramberg–Osgood stress–strain relation were investigated. Fracture mechanics behavior was observed using scanning electron microscopic analysis of the crack shapes and fracture surfaces. Crack propagation was assessed in relation to the actual crack size and the loading level. Interrupted fatigue tests were also carried out to investigate the kinetics of the fatigue evolution of the materials. Microstructural evaluation of the samples was performed using light, scanning and transmission electron microscopy as well as X-ray diffraction, and measurement of dislocations was completed using TEM and XRD. The course of dislocation density in relation to cumulative usage factor was similar for both steels. However, the nature and distribution of dislocations were different in the individual steels and this resulted in different mechanical behaviors. The nature of the fracture surfaces of both steels appeared similar despite differences in dislocation arrangement. The distances between striation lines initially increased with increasing crack length and then became saturated. The low cycle fatigue behavior investigated can provide a reference for the remaining life assessment and lifetime extension analysis of nuclear power plant components.

  17. Lessons learnt from PSA for new and advanced reactors in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Tokmachev, G.; Morozov, V. [JSC ' ' Atomenergoproekt' ' , Moscow (Russian Federation)

    2011-11-15

    Customer requirements to probabilistic safety targets are usually stronger than existing Regulatory or IAEA ones. It appears that industry takes the lead over regulation in this case and forces the designer to find and implement appropriate means to enhance safety, which sometimes have no reference to practical experience. On the other hand, regulatory documents and the existing PSA methodology are mainly oriented to operating plants. This creates problems when developing a PSA as well as performing regulatory reviews. The scope of the PSA may be different depending on a design stage such as the development conceptual, basic or detailed design. In addition, the base case PSA is usually performed for NPP in design. However, a customer may require additional PSA applications to consider, for instance, risk monitoring. In this case the scope of the PSA should be extended to implement special attributes of the application needed that often requires specific information not available at the design stage. Lack of design information affecting PSA development may be associated with incompleteness of the design that is typical for interim design stages and communication problems caused by the involvement of many different companies in the deign activity. To deal with this issue bounding technologies and the iterative PSA development are used. However this sometimes contradicts to the ''best estimate'' approach recommended by regulatory guides. PSA development for advanced NPPs has raised some issues originated from unknown new components, processes and technologies incorporated into the design of an advanced plant. The paper addresses some issues resolved while carrying out PSAs for advanced NPPs. Some PSA results for new advanced VVER plants under construction and the first lessons learnt from the Fukushima accident are also discussed. (orig.)

  18. Confinement matrices for low- and intermediate-level radioactive waste

    Science.gov (United States)

    Laverov, N. P.; Omel'Yanenko, B. I.; Yudintsev, S. V.; Stefanovsky, S. V.

    2012-02-01

    Mining of uranium for nuclear fuel production inevitably leads to the exhaustion of natural uranium resources and an increase in market price of uranium. As an alternative, it is possible to provide nuclear power plants with reprocessed spent nuclear fuel (SNF), which retains 90% of its energy resource. The main obstacle to this solution is related to the formation in the course of the reprocessing of SNF of a large volume of liquid waste, and the necessity to concentrate, solidify, and dispose of this waste. Radioactive waste is classified into three categories: low-, intermediate-, and high-level (LLW, ILW, and HLW); 95, 4.4, and 0.6% of the total waste are LLW, ILW, and HLW, respectively. Despite its small relative volume, the radioactivity of HLW is approximately equal to the combined radioactivity of LLW + ILW (LILW). The main hazard of HLW is related to its extremely high radioactivity, the occurrence of long-living radionuclides, heat release, and the necessity to confine HLW for an effectively unlimited time period. The problems of handling LILW are caused by the enormous volume of such waste. The available technology for LILW confinement is considered, and conclusion is drawn that its concentration, vitrification, and disposal in shallow-seated repositories is a necessary condition of large-scale reprocessing of SNF derived from VVER-1000 reactors. The significantly reduced volume of the vitrified LILW and its very low dissolution rate at low temperatures makes borosilicate glass an ideal confinement matrix for immobilization of LILW. At the same time, the high corrosion rate of the glass matrix at elevated temperatures casts doubt on its efficient use for immobilization of heat-releasing HLW. The higher cost of LILW vitrification compared to cementation and bitumen impregnation is compensated for by reduced expenditure for construction of additional engineering barriers, as well as by substantial decrease in LLW and ILW volume, localization of shallow

  19. Modeling the transport of nitrogen in an NPP-2006 reactor circuit

    Science.gov (United States)

    Stepanov, O. E.; Galkin, I. Yu.; Sledkov, R. M.; Melekh, S. S.; Strebnev, N. A.

    2016-07-01

    Efficient radiation protection of the public and personnel requires detecting an accident-initiating event quickly. Specifically, if a heat-exchange tube in a steam generator is ruptured, the 16N radioactive nitrogen isotope, which contributes to a sharp increase in the steam activity before the turbine, may serve as the signaling component. This isotope is produced in the core coolant and is transported along the circulation circuit. The aim of the present study was to model the transport of 16N in the primary and the secondary circuits of a VVER-1000 reactor facility (RF) under nominal operation conditions. KORSAR/GP and RELAP5/Mod.3.2 codes were used to perform the calculations. Computational models incorporating the major components of the primary and the secondary circuits of an NPP-2006 RF were constructed. These computational models were subjected to cross-verification, and the calculation results were compared to the experimental data on the distribution of the void fraction over the steam generator height. The models were proven to be valid. It was found that the time of nitrogen transport from the core to the heat-exchange tube leak was no longer than 1 s under RF operation at a power level of 100% N nom with all primary circuit pumps activated. The time of nitrogen transport from the leak to the γ-radiation detection unit under the same operating conditions was no longer than 9 s, and the nitrogen concentration in steam was no less than 1.4% (by mass) of its concentration at the reactor outlet. These values were obtained using conservative approaches to estimating the leak flow and the transport time, but the radioactive decay of nitrogen was not taken into account. Further research concerned with the calculation of thermohydraulic processes should be focused on modeling the transport of nitrogen under RF operation with some primary circuit pumps deactivated.

  20. Investigation of the possibility of using residual heat reactor energy

    Science.gov (United States)

    Aminov, R. Z.; Yurin, V. E.; Bessonov, V. N.

    2017-11-01

    The largest contribution to the probable frequency of core damage is blackout events. The main component of the heat capacity at each reactor within a few minutes following a blackout is the heat resulting from the braking of beta-particles and the transfer of gamma-ray energy by the fission fragments and their decay products, which is known as the residual heat. The power of the residual heat changes gradually over a long period of time and for a VVER-1000 reactor is about 15–20 MW of thermal power over 72 hours. Current cooldown systems increase the cost of the basic nuclear power plants (NPP) funds without changing the amount of electricity generated. Such systems remain on standby, accelerating the aging of the equipment and accordingly reducing its reliability. The probability of system failure increases with the duration of idle time. Furthermore, the reactor residual heat energy is not used. A proposed system for cooling nuclear power plants involves the use of residual thermal power to supply the station’s own needs in emergency situations accompanied by a complete blackout. The thermal power of residual heat can be converted to electrical energy through an additional low power steam turbine. In normal mode, the additional steam turbine generates electricity, which makes it possible to ensure spare NPP and a return on the investment in the reservation system. In this work, experimental data obtained from a Balakovo NPP was analyzed to determine the admissibility of cooldown of the reactors through the 2nd circuit over a long time period, while maintaining high-level parameters for the steam generated by the steam generators.

  1. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    Energy Technology Data Exchange (ETDEWEB)

    Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland)

    1997-12-31

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

  2. Results of operation and current safety performance of nuclear facilities located in the Russian Federation

    Science.gov (United States)

    Kuznetsov, V. M.; Khvostova, M. S.

    2016-12-01

    After the NPP radiation accidents in Russia and Japan, a safety statu of Russian nuclear power plants causes concern. A repeated life time extension of power unit reactor plants, designed at the dawn of the nuclear power engineering in the Soviet Union, power augmentation of the plants to 104-109%, operation of power units in a daily power mode in the range of 100-70-100%, the use of untypical for NPP remixed nuclear fuel without a careful study of the results of its application (at least after two operating periods of the research nuclear installations), the aging of operating personnel, and many other management actions of the State Corporation "Rosatom", should attract the attention of the Federal Service for Ecological, Technical and Atomic Supervision (RosTekhNadzor), but this doesn't happen. The paper considers safety issues of nuclear power plants operating in the Russian Federation. The authors collected statistical information on violations in NPP operation over the past 25 years, which shows that even after repeated relaxation over this period of time of safety regulation requirements in nuclear industry and highly expensive NPP modernization, the latter have not become more safe, and the statistics confirms this. At a lower utilization factor high-power pressure-tube reactors RBMK-1000, compared to light water reactors VVER-440 and 1000, have a greater number of violations and that after annual overhauls. A number of direct and root causes of NPP mulfunctions is still high and remains stable for decades. The paper reveals bottlenecks in ensuring nuclear and radiation safety of nuclear facilities. Main outstanding issues on the storage of spent nuclear fuel are defined. Information on emissions and discharges of radioactive substances, as well as fullness of storages of solid and liquid radioactive waste, located at the NPP sites are presented. Russian NPPs stress test results are submitted, as well as data on the coming removal from operation of NPP

  3. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

    The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

  4. Simulation of a loss of coolant accident: Results of a standard problem exercise of the International Atomic Energy Agency

    Energy Technology Data Exchange (ETDEWEB)

    Sloan, S.M.; Hassan, Y. (Texas A M Univ., College Station (USA))

    1989-01-01

    The purpose of this study was to compare the results generated from the IBM version of RELAP5/MOD2 to the experimental data of an International Atomic Energy Agency (IAEA) standard problem exercise. The standard problem exercise data were that of a 7.4% break loss-of-coolant accident conducted at a test facility in Hungary. The United States did not formally participate in this exercise whose aim was to assess the capabilities of computer codes and modeling techniques and in which a total of 17 organizations from 12 countries participated. The results obtained using the IBM version of RELAP5/MOD2 compared favorably with the experimental data. The experimental facility, PMK-NVH (Paks Model Circuit), is a scaled-down model of a Hungarian reactor, the VVER-440 Paks nuclear power plant. A volume and power scaling ratio of 1:2070 is used. The six loops of the actual reactor are modeled by one active loop called the PMK. The secondary loop in the experimental facility is the NVH loop. The coolant in the facility is water, and the operating conditions are the same as in the actual reactor. The orientation of the steam generator is horizontal, as opposed to the vertical design of once-through and U-tube steam generators. The parameters of the accident are that it starts at full power, a 3-mm cold-side break occurs at the upper head of the downcomer, there is no injection from hydroaccumulators, the high-pressure injection system corresponds to the case in which one-third of the pumps are available, and isolation of the secondary occurs immediately after transient initiation.

  5. A Review of Thorium Utilization as an option for Advanced Fuel Cycle--Potential Option for Brazil in the Future

    Energy Technology Data Exchange (ETDEWEB)

    Maiorino, J.R.; Carluccio, T.

    2004-10-03

    Since the beginning of Nuclear Energy Development, Thorium was considered as a potential fuel, mainly due to the potential to produce fissile uranium 233. Several Th/U fuel cycles, using thermal and fast reactors were proposed, such as the Radkwoski once through fuel cycle for PWR and VVER, the thorium fuel cycles for CANDU Reactors, the utilization in Molten Salt Reactors, the utilization of thorium in thermal (AHWR), and fast reactors (FBTR) in India, and more recently in innovative reactors, mainly Accelerator Driven System, in a double strata fuel cycle. All these concepts besides the increase in natural nuclear resources are justified by non proliferation issues (plutonium constrain) and the waste radiological toxicity reduction. The paper intended to summarize these developments, with an emphasis in the Th/U double strata fuel cycle using ADS. Brazil has one of the biggest natural reserves of thorium, estimated in 1.2 millions of tons of ThO{sub 2}, as will be reviewed in this paper, and therefore R&D programs would be of strategically national interest. In fact, in the past there was some projects to utilize Thorium in Reactors, as the ''Instinto/Toruna'' Project, in cooperation with France, to utilize Thorium in Pressurized Heavy Water Reactor, in the mid of sixties to mid of seventies, and the thorium utilization in PWR, in cooperation with German, from 1979-1988. The paper will review these initiatives in Brazil, and will propose to continue in Brazil activities related with Th/U fuel cycle.

  6. Comprehensive investigation of the corrosion state of the heat exchanger tubes of steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Nemeth, Z.; Varga, K.; Baja, B.; Rado, K. [Pannonia Univ., Institute of Radiochemistry and Radioecology, Veszprem (Hungary); Szabo, N.A. [Istvan Szechenyi Univ., Dept. of Physics and Chemistry, Gyor (Hungary); Homonnay, Z.; Kuzmann, E. [Eotvos Lorand Univ., Institute of Chemistry, Budapest (Hungary); Patek, G.; Schunk, J. [Paks NPP Ltd., Paks (Hungary)

    2009-07-01

    Evaluating the water chemistry in the primary circuit and the effect of chemical decontamination of the heat exchanger tubes performed by the AP-CITROX procedure at Paks NPP (Hungary), a project dealing with the comprehensive investigation of the general corrosion state of the steam generators (SGs) has been initiated. Owing to the fact that there is no investigation method available for the in-situ monitoring of the inner surfaces of heat exchanger tubes, a research program based on sampling as well as on ex-situ electrochemical and surface analytical measurements were developed and elaborated. In the time period of 2000-2008 - within the frame of the above project - 45 stainless steel specimens, cut out from various locations of the steam generators of the Paks NPP were investigated. Besides to the corrosion characteristics (corrosion rate, thickness and chemical composition of the protective oxide-layer) surface properties (morphology, chemical and phase compositions) of the passive layer formed on the inner surface of above heat exchanger tubes were studied, too. The passivity of the inside surface of the stainless steel specimens was measured by voltammetry, the morphology, chemical and phase compositions of the oxide layer formed on the surface were analyzed by SEM-EDX, XRD and CEMS methods. The great number of experimental results allowed us to develop an electronic database which involves the results of the above corrosion experiments, and also some special characteristics of the tubes (e.g. location in the SGs, surface pretreatment by decontamination, if any, etc). Evaluating the main relations among these parameters may contribute to the identification of important processes affecting the corrosion state of steam generators, and highly decisive concerning a life time prolongation project of VVER-type nuclear reactors. In the present work we provide a brief overview on these experiments, some characteristic results, the database developed, as well as some

  7. RADIOACTIVE WASTE MANAGEMENT IN THE USSR: A REVIEW OF UNCLASSIFIED SOURCES, 1963-1990

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, D. J.; Schneider, K. J.

    1990-03-01

    The Soviet Union operates a vast and growing radioactive waste management system. Detailed information on this system is rare and a general overall picture only emerges after a review of a great deal of literature. Poor waste management practices and slow implementation of environmental restoration activities have caused a great deal of national concern. The release of information on the cause and extent of an accident involving high-level waste at the Kyshtym production reactor site in 1957, as well as other contamination at the site, serve to highlight past Soviet waste management practices. As a result, the area of waste management is now receiving greater emphasis, and more public disclosures. Little is known about Soviet waste management practices related to uranium mining, conversion, and fuel fabrication processes. However, releases of radioactive material to the environment from uranium mining and milling operations, such as from mill tailings piles, are causing public concern. Official Soviet policy calls for a closed fuel cycle, with reprocessing of power reactor fuel that has been cooled for five years. For power reactors, only VVER-440 reactor fuel has been reprocessed in any significant amount, and a decision on the disposition of RBMK reactor fuel has been postponed indefinitely. Soviet reprocessing efforts are falling behind schedule; thus longer storage times for spent fuel will be required, primarily at multiple reactor stations. Information on reprocessing in the Soviet Union has been severely limited until 1989, when two reprocessing sites were acknowledged by the Soviets. A 400-metric ton (MT) per year reprocessing facility, located at Kyshtym, has been operational since 1949 for reprocessing production reactor fuel. This facility is reported to have been reprocessing VVER-440 and naval reactor fuel since 1978, with about 2000 MT of VVER-440 fuel being reprocessed by July 1989. A second facility, located near Krasnoyarsk and having a 1500 MT per

  8. Influence of the composition on the radiation embrittlement alloys; Einfluss der Zusammensetzung auf die Strahlenversproedungslegierungen

    Energy Technology Data Exchange (ETDEWEB)

    Boehmert, J.; Kryukov, A.; Nikolaev, Yu.A.; Korolev, Yu.N.; Erak, D.Yu.; Gerashenko, S.S.

    1999-02-01

    The radiation embrittlement of the reactor pressure vessel is highly safety-relevant for VVER-type pressure vessels. The sensitivity against radiation embrittlement depends on the chemical composition of the pressure vessel steel. Using an irradiation experiment at surveillance positions in two Russian VVER 440-type reactors the effects of copper, phosphorus and nickel on the radiation embrittlement should be investigated. For that, eight mock-up alloys were selected. Their chemical composition varied between 0.015 and 0.42% Cu, 0.002 and 0.039% P, 0.01 and 1.98% Ni, 0.09 and 0.37% Si, and 0.35 and 0.49% Mn. Charpy-V impact tests and tensile tests were performed with specimens machined from these alloys. The specimens were tested in the as-received state, in the irradiated state (fluence: 1x10{sup 19} and 8x10{sup 19}/cm{sup 2} [E>0.5 MeV]) and in the post-irradiation annealed state. In the as-received state, the alloys have a ferritic microstructure. Apart from Cu, the alloyed elements are solved in the matrix. Irradiation produces strong hardening and embrittlement. The effect increases with the Cu and P content. Ni causes an additional embrittlement. It is independent on the Ni concentration within the range of 1.1 to 2% Ni and results in a shift of the ductile-brittle transition temperature of about 120 C after a fluence of 1x10{sup 19}/cm{sup 2} by a flux of 4x10{sup 11}/cm{sup 2} s. The shift does not depend on the Cu or P content. Furthermore, the upper shelf energy is especially reduced by the Mi-rich alloys. For very low content of Cu and P these relations are not valid. The irradiation effect can be eliminated by annealing at 475 C/100 h. For high content of Cu or P the recovery is incomplete, it remains a residue of 20 to 25% of the irradiation effect. Ni has no influence on the recovery. Comparing the results of this study with the ones of the surveillance programmes of the VVER 440-type reactors, the alloys with low Ni content show the same irradiation

  9. PostgreSQL 8.3

    Directory of Open Access Journals (Sweden)

    Pavel Stěhule

    2007-12-01

    čuje implementací dalších modulů SQL. Ve verzi 8.3 je to konkrétně SQL/XML (rozšíření ANSI SQL, která umožňuje operace s XML dokumenty přímo v databázi a zjednodušuje generování XML dokumentů. Zásadní (interní změnou je zkrácení hlavičky řádku z 28 bajtů na 24 bajtů. Další změnou, která by měla vést k minimalizaci velikosti uložených dat je diverzifikace typu varlena. Tento typ se v PostgreSQL používá pro serializaci hodnot všech typů s variabilní délkou. Trochu připomíná string v Pascalu. První byty nesou informaci o délce, další nesou obsah. Starší verze PostgreSQL znaly jen typ varlena s 4byte informací o délce. 8.3 podporuje také typ varlena s 1byte záhlavím. Úspora by se měla projevit hlavně u typu NUMERIC a krátkých řetězců. K překladu PostgreSQL lze počínaje touto verzí použít jak gcc, MINGW tak Microsoft Visual C++ (na platformě Microsoft Windows.

  10. Visualization of Multidimensional Data in Purpose of Qualitative Classification of Various Types of Coal / Wizualizacja Wielowymiarowych Danych W Celu Klasyfikacji JAKOŚCIOWEJ RÓŻNYCH TYPÓW WĘGLA

    Science.gov (United States)

    Niedoba, Tomasz; Jamróz, Dariusz

    2013-12-01

    ędzy badanymi zmiennymi wektora X. Współczynniki korelacji liniowej są wyznaczane dla par zmiennych losowych całkowicie niezależnie od pozostałych zmiennych. Cząstkowe współczynniki korelacji liniowej wyznaczane są w oparciu o macierz współczynniki korelacji liniowej z uwzględnieniem roli pozostałych zmiennych w rozważanym równaniu regresji liniowej. W przypadku analizy trzech zmiennych losowych, z których jedna jest traktowana jako zmienna zależna a dwie pozostałe jako niezależne sprowadza się to do wyznaczania współczynników korelacji dla zrzutowanych punktów równolegle do płaszczyzny regresji na ściany układu współrzędnych. Pozwala to wyznaczyć hierarchię (siłę wpływu) zależności zmiennych w rozpatrywanym układzie. Na analizie macierzy współczynników korelacji liniowej oparta jest analiza czynnikowa, która pozwala pogrupować występujące zmienne w tzw. czynniki, które reprezentują połączone wpływy zmiennych na rezultaty rozpatrywanych procesów, czyli przeprowadzić pewną klasyfikację zmiennych. W klasyfikacji typów węgli wyróżnia się wiele typów, z umownym podziałem na węgle energetyczne i koksujące. Dane dotyczące węgla są traktowane zwykle jako niezależne wielkości, przy czym takie podejście nie zawsze jest właściwe. Autorzy zaproponowali nowe rozwiązania w tym zakresie i dokonali wielowymiarowej analizy trzech wybranych typów węgla o różnych właściwościach (węgle typu 31, 34.2 oraz 35), które pochodziły z trzech różnych kopalń zlokalizowanych w Górnośląskim Okręgu Przemysłowym. Obiektem badań w każdej z tych kopalń był tzw. węgiel surowy, nie poddawany procesom przeróbczym. Dla każdego z węgli dokonano szczegółowej analizy wybranych siedmiu cech, opisujących jego właściwości, których przykładowe wyniki zostały zaprezentowane w tabelach 1-3. Aby dokonać adekwatnej i dokładnej analizy statystycznej zebranych danych konieczna jest wielowymiarowa analiza wybranych cech

  11. PORÓWNANIE WYBRANYCH WŁAŚCIWOŚCI ZAPRAW ŻYWICZNYCH ZAWIERAJĄCYCH ODPADOWE TWORZYWA SZTUCZNE

    Directory of Open Access Journals (Sweden)

    Bernardeta DĘBSKA

    Full Text Available Racjonalna gospodarka odpadami stanowi jeden z priorytetowych kierunków szeroko rozumianej ochrony środowiska. Fakt umiejętnego zagospodarowania odpadów jest także ważny w kontekście zrównoważonego rozwoju społeczeństw. Do odpadów wyjątkowo uciążliwych dla środowiska zaliczyć należy tworzywa sztuczne. Wzrastające nieustannie ilości tego typu odpadów powodują występowanie problemów zarówno ekologicznych, jak i gospodarczych, co związane jest ze słabą biodegradacją tworzyw. Odpady te zaczęto wykorzystywać do produkcji materiałów budowlanych. Badania nad zagospodarowaniem odpadów z tworzyw sztucznych prowadzone są obecnie w różnych ośrodkach naukowych na świecie. W literaturze można znaleźć opisy wykorzystania odpadów m.in. polietylenu i polipropylenu, styropianu, poliuretanów, poliwęglanu, poliamidu, czy poli(chlorku winylu, jako modyfikatorów betonów i zapraw cementowych. W niniejszym artykule przedstawiono wyniki porównania wybranych właściwości czterech serii zapraw żywicznych zawierających różne odpady tworzyw sztucznych tj.: polipropylen (PP, polietylen (PE, piankę poliuretanową (PU oraz ekspandowany polistyren (EPS. Odpady te pochodziły odpowiednio z kubków po jogurtach, pianki podkładowej pod panele, pianki montażowej oraz płyt styropianowych. Zostały one rozdrobnione i stanowiły częściowy zamiennik kruszywa w zaprawach epoksydowych. Zbadano takie właściwości zapraw, jak: wytrzymałość na zginanie i ściskanie, gęstość objętościowa oraz nasiąkliwość. Wskazano materiał odpadowy, umożliwiający otrzymanie zaprawy cechującej się najkorzystniejszymi wartościami oznaczonych parametrów. Na podstawie uzyskanych wyników badań stwierdzono, że nawet przy 20% substytucji piasku odpadami tworzyw sztucznych, można otrzymać kompozyt charakteryzujący się bardzo dobrymi parametrami wytrzymałościowymi oraz niską nasiąkliwością wodą.

  12. Correction of location of boundaries in cadastre modernization process

    Science.gov (United States)

    Hanus, Paweł

    2013-06-01

    ępowania pozwoliłby uniknąć wielu nieporozumień i trudności przy wykorzystywaniu danych zgromadzonych w ewidencji gruntów i budynków, a z drugiej strony nie zaburzałby dotychczas stosowanych procedur, których celem jest ustalenie granic z dokładnością właściwą dla tego typu szczegółów terenowych.

  13. Wpływ leków przeciwpsychotycznych na występowanie zespołu metabolicznego

    Directory of Open Access Journals (Sweden)

    Adam Wysokiński

    2014-12-01

    Full Text Available Zespół metaboliczny to zbiór współwystępujących zaburzeń, które zwiększają ryzyko zawału serca, udaru i cukrzycy. W Europie do oceny zespołu metabolicznego stosuje się kryteria International Diabetes Federation. Zgodnie z nimi zespół metaboliczny to współobecność otyłości brzusznej (trzewnej, centralnej, zdefiniowanej jako obwód talii ≥80 cm u Europejek i ≥94 cm u Europejczyków, oraz dwóch spośród następujących czynników: 1 stężenia trójglicerydów >1,7 mmol/l (150 mg/dl lub leczenia triglicerydemii; 2 stężenia cholesterolu frakcji HDL <1,0 mmol/l (40 mg/dl u mężczyzn i <1,3 mmol/l (50 mg/dl u kobiet lub leczenia tego zaburzenia lipidowego; 3 ciśnienia tętniczego skurczowego ≥130 mm Hg bądź rozkurczowego ≥85 mm Hg lub leczenia wcześniej rozpoznanego nadciśnienia; 4 stężenia glukozy w osoczu na czczo ≥5,6 mmol/l (100 mg/dl lub rozpoznanej wcześniej cukrzycy typu 2. Zespół metaboliczny występuje 2–3-krotnie częściej u pacjentów otrzymujących leki przeciwpsychotyczne. Obecność zespołu metabolicznego wiąże się z 2–3-krotnie większym ryzykiem zgonu z powodu powikłań sercowo-naczyniowych. Niemal wszystkie leki przeciwpsychotyczne podnoszą ryzyko wystąpienia zespołu metabolicznego; dotyczy to w szczególności leków najskuteczniej działających (klozapina, olanzapina. Punktem wyjścia metabolicznych powikłań leczenia przeciwpsychotycznego jest wzrost łaknienia indukowany leczeniem, prowadzący do przyrostu masy ciała; powikłania te mogą się rozwijać wskutek bezpośredniego działania leków. Zdecydowana większość pacjentów nie otrzymuje właściwego leczenia zaburzeń metabolicznych. W przypadku polekowego przyrostu masy ciała dostępne opcje terapeutyczne są mało skuteczne i zbyt rzadko stosowane.

  14. OCHRONA ŚRODOWISKA W TURYSTYCE NA PODKARPACIU

    Directory of Open Access Journals (Sweden)

    Galina KALDA

    2016-03-01

    Full Text Available W artykule prezentowano analizę środowiska województwa podkarpackiego zanieczyszczanego odpadami turystycznymi. Województwo podkarpackie jest popularne z miejsc aktywności turystycznej oraz dobrej sieci infrastruktury turystycznej. Jak pokazują badania, Podkarpacie jest województwem z nieznacznym zanieczyszczeniem środowiska. Wody powierzchniowe są głównym źródłem zapotrzebowania na Podkarpaciu. Poprzez swoje położenie geograficzne województwo można uznać za jedno z niewielu miejsc z Polsce, gdzie sezon turystyczny trwa praktycznie cały rok. Podkarpacie przez to przyciąga turystów, którzy pragną aktywnego odpoczynku przez obcowanie z przyrodą. Nagły rozwój turystyki został spowodowany promowaniem województwa jako regionu atrakcyjnego turystycznie i przychylnego turystom. Jednym z większych problemów hamujący rozwój turystyki jest zbyt mała sieć informacji turystycznej. Rozmieszczone są one w mało strategicznych dla turystów miejscach. Znaleźć je można nawet kilka kilometrów od przystanków autobusowych czy dworcach kolejowych. Dla turystów, którzy przyjeżdżają do nowego miejsca, ważne jest uzyskanie jakichkolwiek informacji o mieście, noclegach czy zabytkach, które znajdują się w okolicy. Analiza pokazała, że turyści wytwarzają średnio do 30 razy więcej odpadów, niż przeciętny mieszkaniec województwa podkarpackiego, co spowodowane jest wykorzystywaniem większej ilości opakowań jednorazowych, wytwarzaniem odpadów, takich jak resztki żywności, tekstylia, szkło itp. Po przeprowadzonej analizie stwierdzono, ze turystyka pośrednio wpływa na zanieczyszczenia gleby, powierza oraz ma nieznaczny udział w poziomie hałasu komunikacyjnego. Nie oznacza to jednak, że jej wpływ jest niegroźny i nieistotny. Aby o tym się przekonać należy wykonać specjalne badania, które określiłyby stopień, w jakim turystyka może przyczynia się do tego typu zanieczyszczeń.

  15. Wybrane farmakokinetyczne interakcje leków w trakcie leczenia padaczki. Część I

    Directory of Open Access Journals (Sweden)

    Magdalena Justyna Kacperska

    2013-04-01

    Full Text Available Padaczka jest jedną z najdłużej znanych chorób. Słowo epilepsia liczy 2500 lat i pochodzi od greckiego epilamvanein, co znaczy ‘atakować’, ‘chwycić’, ‘posiąść’. Napady padaczkowe traktowane były jako wyraz owładnięcia przez demony, złe duchy, w związku z czym padaczkę przez długi czas uważano za „świętą chorobę”. Nie jest to choroba w klasycznym znaczeniu, a raczej skomplikowany proces patofizjologiczny, którego bardzo liczne i złożone objawy są wynikiem różnych zaburzeń funkcji mózgu. Padaczka należy do najtrudniejszych problemów neuroepidemiologicznych. Napady padaczkowe są wyrazem patologicznej czynności różnych obszarów mózgu w przebiegu wielu procesów chorobowych. Źródłem patologicznych wyładowań w klinicznej formie napadu padaczkowego mogą być blizny pourazowe, zmiany uciskowe, zapalne, zwyrodnieniowe, ogniska naczyniopochodne czy zaburzenia rozwojowe. Ogniskiem padaczkowym jest strefa zmienionej tkanki, leżącej między uszkodzeniem a okolicą zdrową. To grupa neuronów generująca okresowo napadową czynność bioelektryczną w formie napadowych wy- ładowań depolaryzacyjnych generujących kliniczny napad padaczkowy. Większość padaczek to zaburzenia pierwotne mózgowe, choć istnieje również wiele procesów pozamózgowych zaburzających homeostazę ustrojową. W leczeniu padaczki nie występuje jeden standardowy sposób postępowania. Celem terapii jest całkowita kontrola napadów i uzyskanie jak najmniejszych objawów niepożądanych podczas leczenia lekami przeciwpadaczkowymi. Wiedza i doświadczenie lekarzy praktyków są najistotniejszym czynnikiem wpływającym na opiekę nad chorym z padaczką. Lek powinien być dostosowany do typu napadu lub zespołu padaczkowego, częstości i ciężkości napadów. Pojawienie się leków nowej generacji dało im pewną przewagę w stosunku do starszych leków. Cechują je: większa swoistość działania, lepsze w

  16. Wybrane farmakokinetyczne interakcje leków w trakcie leczenia padaczki. Część II

    Directory of Open Access Journals (Sweden)

    Karol Jastrzębski

    2013-04-01

    Full Text Available Padaczka to choroba o nieznanej do końca etiologii, charakteryzująca się występowaniem nieprowokowanych napadów padaczkowych. Napad padaczkowy to z kolei przejściowa zmiana reaktywności lub zmiana stanu fizjologicznego części bądź całego mózgu. Napady dzielą się na: częściowe, uogólnione i niesklasyfikowane. Pojęcie padaczki lekoopornej może się wydawać oczywiste i zrozumiałe, niemniej jednak nie opracowano dotychczas powszechnie uznawanej szczegółowej definicji. W efekcie lekarze i badacze stosują bardzo różne kryteria, a w niektó- rych przypadkach nawet rezygnują z dokładnych kryteriów, co znacznie utrudnia porównywanie wyników badań klinicznych i tworzenie wytycznych. W leczeniu padaczki nie występuje jeden standardowy sposób postępowania. Celem terapii padaczki jest całkowita kontrola napadów i uzyskanie jak najmniejszych objawów niepożądanych podczas leczenia lekami przeciwpadaczkowymi. Lek powinien być dostosowany do typu napadu lub zespołu padaczkowego, częstości i ciężkości napadów. Wybór leków zależy od rodzaju napadów, przykładowo w napadach pierwotnych uogólnionych stosowany jest kwas walproinowy, natomiast we wtórnie uogólnionych i częściowych – karbamazepina. Leki starszej generacji (fenytoina, fenobarbital, prymidon powoli wychodzą z użycia. Mogą być jednak przepisywane z powodu indywidualnych wskazań. Jest też bardzo duża grupa nowych leków (lamotrygina, wigabatryna, okskarbazepina, gabapentyna, lewetyracetam, felbamat, topiramat, tiagabina, które stają się coraz bardziej popularne. Pojawienie się leków nowej generacji dało im pewną przewagę w stosunku do starszych leków. Cechują je: większa swoistość działania, lepsze właściwości farmakokinetyczne, lepsza ocena klinicznych prób i słabsze objawy niepożądane. Z badań klinicznych i z bezpośrednich obserwacji wynika, iż są to leki bardzo przydatne w niektórych typach padaczek. Nie ulega w

  17. Srovnání hodnocení chování rodičů a trenérů u tenistů a tenistek různých věkových skupin Comparison of assessments of parents' and coaches' behaviour by male and female tennis players of different ages

    Directory of Open Access Journals (Sweden)

    Tjaša Filipčič

    2006-02-01

    Full Text Available Tenis je určitě jednou z těch sportovních her, v nichž je úspěch ovlivňován psychologickými schopnostmi. Během celé své sportovní dráhy jsou obvykle tenisoví hráči vystaveni psychologickému tlaku. Cílem této studie bylo zjistit, jak mladí tenisoví hráči hodnotí chování (psychologický tlak a aktivitu své matky, otce a tenisového trenéra. Vzorek zahrnoval 96 tenistů a 96 tenistek náležejících do třech věkových kategorií, kteří vyplňovali tři dotazníky uzavřeného typu. Údaje byly zpracovávány v souladu s výzkumnými cíly. Pro všechny proměnné byly podle pohlaví vypočítávány popisné statistické parametry: střední hodnota, standardní odchylka, minimum, maximum, šikmost, špičatost a Kolmogorovův-Smirnovův test normality. Pro srovnávání hodnocení chování hráčova otce, matky a trenéra podle pohlaví a posléze pro srovnávání tří věkových kategorií (U12, U14, U16 byla použita analýza variance (ANOVA. Srovnávání hodnocení chování otce, matky a trenéra podle pohlaví odhalilo statisticky významné rozdíly pouze u hodnocení otce. Při srovnávání hodnocení chování obou rodičů a trenéra podle věkových kategorií byly statisticky významné rozdíly stanoveny opět pouze u hodnocení otce. Tennis is clearly one of those sports games where success is influenced by one's psychological abilities. Throughout their sports careers tennis players are usually exposed to psychological pressure. The aim of the study was to investigate how young tennis players assess the behaviour (psychological pressure and activity of their mother, father and tennis coach. The sample included 96 male and 96 female players distributed in three age categories who filled in three questionnaires of the closed type. Data were processed in accordance with the research goals. Descriptive statistics parameters were calculated for all variables by gender: mean value, standard deviation

  18. On “The dictionary of active Polish and Ukrainian phraseology” [Leksykon aktywnej frazeologii polskiej i ukraińskiej]. Contrastive linguistics and culture

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    Wojciech Paweł Sosnowski

    2017-12-01

    Full Text Available On The dictionary of active Polish and Ukrainian phraseology [Leksykon aktywnej frazeologii polskiej i ukraińskiej]. Contrastive linguistics and culture The Dictionary of Active Polish and Ukrainian Phraseology [Leksykon aktywnej frazeologii polskiej i ukraińskiej] is the first publication of its kind in the history of Polish and Ukrainian lexicography. It consists of equivalent phrasal units in Polish and Ukrainian. The innovative aspect of the lexicon is that it uses a semantic metalanguage to establish equivalent units. The authors developed a new method of searching for equivalent units which uses the meaning — not the form — as the starting point. This method enables the identification of equivalent units in both languages. Moreover, it enables the identification of units that do not have equivalents. The units which lack equivalents are usually deeply rooted in Poland’s or Ukraine’s historical and cultural context, and are thus defined as culturemes. Even though they lack equivalents, it was decided not to exclude them from the Leksykon’s structure, as they are actively used by the speakers of Polish and Ukrainian. This paper provides an overview of the Leksykon’s methodology and presents the authors’ definition of phraseologism. The most important points in the paper are illustrated with a number of example entries from the dictionary. The primary focus of the paper rests on phrasal units which lack equivalents.   O Leksykonie aktywnej frazeologii polskiej i ukraińskiej. Konfrontacja językowa a kultura Opracowywany przez nas Leksykon aktywnej frazeologii polskiej i ukraińskiej jest pierwszym dziełem tego typu w historii leksykografii polskiej i ukraińskiej. W leksykonie prezentujemy odpowiedniości jednostek frazeologicznych w języku polskim i ukraińskim za pomocą semantycznego języka pośrednika. Wyznaczenie kierunku od znaczenia ku formie pozwoliło dobrać ekwiwalenty jednostek frazeologicznych w obu j

  19. Wioski tematyczne w powiecie tucholskim = Theme villages in powiat tucholski

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    Kawska Paulina

    2016-11-01

    tematycznych funkcjonujących w świadomości mieszkańców powiatu tucholskiego, a w szczególności z kojarzeniem terminu wioska tematyczna i wiedzy badanych osób o wioskach tego typu. Badania pokazały również stopień wypromowania produktu turystycznego jakim są wioski tematyczne wśród mieszkańców powiatu tucholskiego oraz ocenę ich atrakcyjności. Abstract The article presents the issue of the theme villages by referring them to the awareness of the Powiat Tucholski residents connected with the villagesexisting in its area. That is why it was decided to conduct a survey. At the verybeginning, the respondents were asked what they associate the term theme village with, further questions were focused on the inhabitant’s knowledge of the set up in the research areavillages. The survey was supposed to show how much the tourist product - theme villages created by the local societies has been promoted among the Powiat Tucholski residents and how attractive it is for them.

  20. WYSTĘPOWANIE ROŚLIN INWAZYJNYCH W OBRĘBIE BUDOWLI I POWIERZCHNI UTWARDZONYCH W DOLINACH RZECZNYCH KARPAT I KOTLINY SANDOMIERSKIEJ

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    Dominik WRÓBEL

    Full Text Available Badania terenowe, prowadzone w latach 2010-2016, w dolinach rzecznych polskiej części Karpat oraz w Kotlinie Sandomierskiej i w przylegającym do niej odcinku doliny Wisły, miały za zadanie uzupełnić, wiedzę o występowaniu inwazyjnych gatunków roślin (inwaderów w najsilniej przekształconych dolinach rzecznych, a w szczególności określić typy zabudowy dolin rzecznych, sprzyjające rozprzestrzenianiu się tych gatunków. Przeanalizowano 118 transektów zlokalizowanych zarówno w regionach górskich, podgórskich i nizinnych, w odcinkach uregulowanych jak i nieuregulowanych dolin rzecznych, cieków o różnej wielkości. Wyodrębniono główne typy/kategorie zabudowy, łączące w sobie: obiekty hydrotechniczne i przeciwpowodziowe, w tym obwałowania, umocnienia brzegowe i ostrogi korytowe (I, mieszkalną i usługową zabudowę śródmiejską (II, drogowe i kolejowe linie komunikacyjne, w tym mosty (III, wyrobiska górnicze, zabudowę produkcyjną, wydobywczą, magazynową i towarzyszącą (IV, zabudowę rozproszoną, ogródki działkowe (V oraz odrębne place, parkingi i składowiska (VI. Na częściach transektów, obejmujących różne formy zabudowy, najczęściej zanotowano występowanie Solidago gigantea / S. canadensis (46, Impatiens glandulifera (30, Echinocystis lobata (22, Robinia pseudoacacia (17, Helianthus tuberosus (15 i Impatiens parviflora (15. Największa liczba stanowisk gatunków inwazyjnych w relacji do wszystkich ich stwierdzeń została zanotowana na różnego rodzaju budowlach hydrotechnicznych, w tym na umocnieniach brzegowych różnego typu. Obserwacje prowadzone w zakresie wpływu inwestycji regulacyjnych na szatę roślinną wskazują, że nie ma istotnych różnic co do zastosowanych sposobów zabudowy umocnieniowej brzegów, które można byłoby uznać za bardziej przyjazne środowisku. W każdym przypadku następuje pozostawianie odkrytego podłoża i promowanie wkraczania inwaderów.

  1. Flexibility in Management of Modernization in Construction - Electrical Works/ Elastyczność W Zarządzaniu Modernizacją Obiektów Budowlanych Na Przykładzie Robót Elektrycznych

    Science.gov (United States)

    Nowotarski, Piotr; Pasławski, Jerzy

    2015-06-01

    wystarczający do pokonania całej drogi na korytarzu). Osiągnięcie kompromisu w tego typu konfliktach interesów może być rozwiązane za pomocą podejścia elastycznego i zwinnego.

  2. Evolution of the Structure and Mechanical Strength of a Coal Particle During Combustion in the Atmosphere of Air and the Mixture of Oxygen and Carbon Dioxide / Ewolucja Struktury Oraz Wytrzymałości Mechanicznej Ziarna Węgla Podczas Spalania W Atmosferze Powietrza Oraz Mieszaninie Tlenu I Dwutlenku Węgla

    Science.gov (United States)

    Pełka, Piotr; Golański, Grzegorz; Wieczorek, Paweł

    2013-09-01

    The research was conducted on the basis of four different types of hard coal and one type of brown coal. There are typical coals commonly used as fuel in Polish CFB boilers. The combustion process was conducted at a temperature of 850°C and the atmosphere of ambient air as well as in the mixture of oxygen and carbon dioxide in different proportions. The research was carried out using specially prepared cubical coal particles with measurements of 15×15mm and also 10×10 mm. The change of the mechanical properties was analyzed based on three parameters, i.e. compression strength, Vickers hardness and fracture toughness. The analysis was supplemented by microscopic images of the surface of the particles using an atomic force microscope. The results obtained clearly indicated the mechanical changes of the coal during its combustion, particularly at the moment of ignition of the char. Moreover, the results correlate very well with the processes of coal comminution that have been described by other authors (Basu, 1999; Chirone et al., 1991) during combustion in the circulating fluidized bed and also explain the sudden change of susceptibility to erosion in the conditions with and without combustion. The measured values can be used as strength parameters in the modelling of the mass loss of coal particles in conditions of circulating fluidized bed combustor that are hard to describe. Badania przeprowadzono na podstawie czterech węgli kamiennych różnego typu oraz jednego węgla brunatnego. Są to typowe węgle energetyczne wykorzystywane powszechnie jako paliwo w kotłach fluidyzacyjnych w Polsce. Proces spalania był prowadzony w temperaturze 850°C w atmosferze powietrza atmosferycznego oraz w atmosferze mieszaniny tlenu oraz dwutlenku węgla w różnych proporcjach. Badania przeprowadzono na spreparowanych do tego celu sześciennych próbkach węgla o wymiarach 15×15 mm oraz 10×10 mm. Zmianę własności mechanicznych przeanalizowano w oparciu o trzy parametry

  3. Wrodzony rodzinnie występujący zespół wydłużonego QT – trudności diagnostyczne

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    Katarzyna Bieganowska

    2009-06-01

    Full Text Available Wrodzony zespół wydłużonego QT (long QT syndrome, LQTS jest uwarunkowaną genetycznie chorobą charakteryzującą się istotnym wydłużeniem odstępu QT oraz nieprawidłowymi załamkami T w zapisie elektrokardiograficznym. Wydłużenie odstępu QT (odzwierciedla wydłużenie repolaryzacji komór sprzyja zasłabnięciom, omdleniom lub nagłej śmierci sercowej w wyniku typowego częstoskurczu komorowego torsade de pointes lub migotania komór. W 1957 roku jako pierwszy został opisany zespół wydłużonego QT z głuchotą – zespół Jervella i Lange-Nielsena – który występuje bardzo rzadko. Niedługo później opisano występujący z częstością ok. 1:2500-5000 zespół Romano-Warda bez głuchoty. Zespół jest genetycznie zróżnicowany, spowodowany mutacjami genów kodujących białka sercowych kanałów jonowych. Najczęściej występujący typ LQT1 zespołu wydłużonego QT (powyżej 50% przypadków jest wynikiem mutacji genu KCNQ1 kodującego białko kanału potasowego IKs, co powoduje nieprawidłową jego funkcję. Badania molekularne są ważne dla ustalenia mutacji. Rozpoznanie zespołu wydłużonego QT bywa trudne, niekiedy konieczne jest wykonanie wielu zapisów EKG, aby udokumentować wydłużenie odstępu QT. Klinicznie wciąż przydatne mogą być kryteria Schwartza i Mossa. Leczenie beta-blokerami jest skuteczne i powinno być wdrażane z chwilą rozpoznania. Należy przede wszystkim unikać wysiłku, stresu oraz leków mogących wydłużyć skorygowany odstęp QT. Niezbędne jest przebadanie rodzin pacjentów z rozpoznanym zespołem wydłużonego QT. W pracy prezentujemy udokumentowany genetycznie przypadek rodzinnie występującego zespołu wydłużonego QT typu 1 (LQT1. Przedstawione zostały problemy diagnostyczne, przebieg kliniczny i wdrożone postępowanie lecznicze.

  4. Pamięć rodu Działowskich

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    Stanisław Roszak

    2011-12-01

    Full Text Available Żyjący do XX w. w ziemi chełmińskiej Działowscy należeli do tych przedstawicieli szlachty i ziemiaństwa, którzy przywiązywali istotne znaczenie do poznawania, budowania i podtrzymywania pamięci o członkach własnej rodziny. Już w I poł. XVIII w. pisarz ziem pruskich Wawrzyniec Działowski na podstawie kwerend w archiwach sądowych i prywatnych z obszaru Prus Królewskich stworzył tzw. czarną księgę Działowskich – zaginiony obecnie rękopis, który zawierał wypisy z akt i pamiętników z XVI-XVIII w. dotyczących tej rodziny oraz rodów z nią skoligaconych. Obok zwykłej ciekawości motywacją pracy Wawrzyńca nad rodzinną genealogią była zapewne chęć wzmocnienia prestiżu i społecznej pozycji rodziny. Stąd też, gdy w wydanym w 1738 r. drugim tomie herbarza Kacpra Niesieckiego pojawiła się bardzo lakoniczna wzmianka o Działowskich, Wawrzyniec, szukając możliwości publicznego przedstawienia znanej sobie przeszłości rodu, interweniował u Józefa Andrzeja Załuskiego a następnie przekazał sumaryczne opracowanie o Działowskich autorowi herbarza, który zamieścił je w suplemencie do czwartego tomu swego dzieła. Z zawartych w nim treści wynika m.in., że już w XVIII w. przedstawiciele tej rodziny rozpowszechniali niezgodną z prawdą legendę o swoim pochodzeniu od średniowiecznego rycerza Mikołaja Działowskiego, znanego z wystąpień po stronie polskiej w czasie wojny trzynastoletniej. Świadectwem troski Działowskich o rodzinną pamięć są także rękopiśmienne zapiski na egzemplarzu wspomnianego tomu przechowywanym w zbiorach Książnicy Kopernikańskiej w Toruniu, a należącym do końca XIX w. do potomków Wawrzyńca. Jego syn, Teodor, na kartach wyklejki odnotowywał w latach 1763-1778 narodzenia, śluby i zgony członków swojej rodziny tworząc rodzaj rodzinnej kroniki linii Działowskich rezydującej w Mgowie i Turznie. Zapiski te, wywodzące się z tradycji rękopiśmiennych ksiąg typu

  5. Regresja zmian skórnych w przebiegu stwardnienia guzowatego u chorej po przeszczepieniu nerki leczonej rapamycyną – opis przypadku i przegląd piśmiennictwa

    Directory of Open Access Journals (Sweden)

    Joanna Renczyńska-Matysko

    2011-04-01

    Full Text Available Wprowadzenie: Stwardnienie guzowate (ang. tuberous sclerosis complex– TSC, choroba Bourneville’a-Pringle’a jest rzadką chorobą należącądo grupy fakomatoz, charakteryzującą się zajęciem skóry, układu nerwowego,narządu wzroku i narządów wewnętrznych. Choroba ta jestzaburzeniem genetycznym powodowanym przez mutacje genów TSC1i TSC2. Produkty tych genów, hamartyna i tuberyna, tworzą komplekshamujący białko mTOR pełniące kluczową rolę w kontroli cyklukomórkowego. Mutacje w genach TSC prowadzą do ciągłej aktywacjiszlaku mTOR, powodując niekontrolowaną proliferację, różnicowanieoraz migrację komórek, czego konsekwencją jest powstawanie malformacjiw wielu narządach. Istotne dla diagnozy stwardnienia guzowategojest badanie dermatologiczne, ponieważ istnieje możliwość rozpoznaniachoroby Bourneville’a-Pringle’a jedynie na podstawie objawówskórnych, które należą do jej kryteriów diagnostycznych. Plamy hipopigmentacyjne,naczyniakowłókniaki, płaskie włókniaki na czole,włókniaki okołopaznokciowe, skóra szagrynowa, znamiona bezbarwnetypu „confetti”, ubytki w szkliwie zębów, włókniaki dziąseł – to najczęstszeobjawy stwardnienia guzowatego. Cel pracy: Przedstawienie regresji zmian skórnych u chorej na stwardnienieguzowate po zastosowaniu rapamycyny jako elementu immunosupresjipo transplantacji nerki. Opis przypadku: Chora 49-letnia ze stwardnieniem guzowatym rozpoznanymw 1988 r. zgłosiła się na Oddział Dermatologii w styczniu2010 r. Pacjentka miała następujące objawy skórne charakterystycznedla TSC: naczyniakowłókniaki, skórę szagrynową, plamy hipopigmentacyjne,guzki Koenena i zmiany skórne typu „confetti”. U kobiety jużw niemowlęctwie obserwowano napady padaczkowe. W 1993 rokuw badaniu metodą tomografii komputerowej (CT zdiagnozowanoobustronne naczyniakotłuszczaki nerek, które doprowadziły doobustronnej niewydolności nerek, a w konsekwencji do

  6. Fingolimod w leczeniu stwardnienia rozsianego

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    Marek Juszczak

    2010-09-01

    Full Text Available W chwili obecnej leczenie immunomodulujące stwardnienia rozsianego (łac. sclerosis multiplex, SM polega na podawaniu leków w postaci iniekcji podskórnych, domięśniowych bądź dożylnych. Powoduje to liczne niedogodności dla pacjentów, wobec czego istnieje zapotrzebowanie na wprowadzenie leku doustnego o skuteczności przewyższającej skuteczność leków obecnie stosowanych. Pierwszym tego typu lekiem jest fingolimod (FTY720. Lek ten dobrze wchłania się z przewodu pokarmowego, a maksymalne stężenie we krwi osiąga po 12-16 godzinach od podania. Mechanizm działania fingolimodu polega na powodowaniu sekwestracji dojrzałych limfocytów w węzłach chłonnych i kępkach Peyera, co zmniejsza ich liczbę we krwi oraz w naciekach limfocytarnych. FTY720 nie upośledza funkcji limfocytów, a zwłaszcza aktywacji limfocytów T. Uważa się, że fingolimod może hamować migrację limfocytów na drodze dwóch niezależnych mechanizmów: pierwszy to zmniejszanie liczby receptorów S1P1 na limfocytach T i osłabienie sygnału do ich migracji, drugi to stałe pobudzanie tych receptorów obecnych na śródbłonku zatok w węzłach chłonnych do wzmacniania bariery ograniczającej migrację limfocytów. Zachęcające wyniki badań klinicznych II fazy umożliwiły zaprojektowanie dwóch dużych badań III fazy, które otworzyły fingolimodowi drogę do rejestracji: FREEDOMS (FTY720 Research Evaluating Effects of Daily Oral therapy in Multiple Sclerosis i TRANSFORMS (Trial Assessing Injectable Interferon versus FTY720 Oral in Relapsing-Remitting Multiple Sclerosis. W zakresie wszystkich klinicznych i rezonansowych punktów końcowych w badaniu FREEDOMS udowodniono przewagę fingolimodu nad placebo. Badanie TRANSFORMS pozwoliło udowodnić, że fingolimod zmniejsza aktywność SM skuteczniej niż interferon β-1a. Dane te potwierdzają, że fingolimod jest obiecują- cym nowym lekiem w terapii SM.

  7. Dopalacze – nowe zagrożenie zdrowia młodzieży

    Directory of Open Access Journals (Sweden)

    Karolina Kierus

    2011-12-01

    Full Text Available Pomimo nowelizacji Ustawy o przeciwdziałaniu narkomanii dopalacze nie zniknęły całkowicie z polskiego rynku i życia młodzieży, gdyż nadal dostępne są w sprzedaży wysyłkowej. Obecnie Internet proponuje młodemu człowiekowi przynajmniej osiem polskojęzycznych stron e-sklepów, oferujących tanią wysyłkę zamówionych towarów do Polski. Skład dopalaczy nie jest w pełni poznany, gdyż podlega stałej ewolucji. Postać, pod jaką występują, to najczęściej susze roślinne (typu Spice, tabletki (party pills, proszki, sprzedawane jako produkty kolekcjonerskie, nienadające się do spożycia przez ludzi. Zażywanie ich polega na paleniu w skrętach, lufkach, fajkach wodnych, przyjmowaniu doustnym oraz w inhalacjach donosowych. Zawierają one zarówno syntetyczne substancje chemiczne, jak i susze bądź sproszkowane nasiona roślin znanych z silnych działań psychostymulujących czy halucynogennych. Efekt działania dopalaczy imituje skutki zażycia popularnych narkotyków, takich jak: marihuana, opium, amfetamina czy ecstasy. Przy długotrwałym stosowaniu dopalacze, podobnie jak inne środki psychostymulujące, mogą powodować uzależnienie, a także być wstępem do eksperymentowania z tzw. twardymi narkotykami. Oprócz działań pobudzających, relaksujących, halucynogennych czy empatogennych powodują wiele działań niepożądanych zarówno z grupy zaburzeń psychicznych, jak też istotnych zaburzeń somatycznych, prowadzących nawet do poważnych zaburzeń świadomości, zaburzeń oddychania czy napadów drgawkowych bezpośrednio zagrażających życiu. Rozpoznanie kliniczne zatrucia dopalaczami ustala się na podstawie objawów klinicznych oraz wywiadu wskazującego na zażycie nieznanej substancji. Z uwagi na brak swoistych odtrutek leczenie pozostaje jedynie objawowe, ma podtrzymywać podstawowe czynności życiowe. Wskazane jest przeprowadzenie konsultacji psychologicznej i/lub psychiatrycznej, a w uzasadnionych

  8. Analysis of Physical Education Students’ Emotional Stability and Reactibility

    Directory of Open Access Journals (Sweden)

    Radka Peřinová

    2015-03-01

    Full Text Available Analysis of Physical Education Students’ Emotional Stability and Reactibility This paper will aim to show the possible association between emotional stability and reaction time variability of Physical Education students. It can be stated that our study confirmed our suppositions which were based on works that have focused on similar topics. Our research sample showed the expected characteristics: primarily lower neuroticism values and higher extraversion when compared to the non-sporting population. Emotional stability which was reflected in the neuroticism dimension in EPQ-R (Eysenck Personality Questionnaire was shown to be connected with variability of the reaction time in the test of reactability to selected visual stimulus, disregarding the reaction rate. The effect of extraversion is partly reflected by the tendency of the sanguine temperament type to react in a balanced manner (i.e. with low reaction time variability during the reactability test. Due to the relatively low number of other temperament types in our sample, it is not possible to draw any conclusions in this regard. Analýza emocionální stability a reaktibility studentů tělesné výchovy Tento příspěvek poukazuje na možnou asociaci mezi emocionální stabilitou a časovou variabilitou dob reakcí u studentů tělesné výchovy. Lze konstatovat, že studie potvrdila naše předpoklady vycházející z odborných prací na obdobná témata. Výzkumný soubor vykazoval předpokládané charakteristiky, především nižších hodnot neuroticismu a vyšší extroverze oproti nesportující populaci. Emocionální stabilita vyjádřená pomocí dimenze neuroticismu (v EPQ-R se ukázala v asociaci s časovou variabilitou dob reakcí v testu reaktibility na výběrový zrakový podnět bez ohledu na rychlost reakce. Vliv extroverze do jisté míry odráží naznačená tendence sangvinického typu temperamentu reagovat vyrovnaně (tedy s nízkou časovou variabilitou dob

  9. Economics of Nuclear Power Plant and the development of nuclear power in Viet Nam

    Energy Technology Data Exchange (ETDEWEB)

    Thanh, Thuy Nguyen Thi; Song, JinHo [University of Science and Technology, Daejeon (Korea, Republic of); Ha, Kwang Soon [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    There are many factors affecting the capital costs like: increased plant size, multiple unit construction, improved construct methods, increase the lifetime of plant and so on, and beside is technical to enhancing the safety for NPPs. For the question that whether building a NPP is really economic than other energy resources or not, we will find the answer by comparing the USD per kWh of different energy sources as: nuclear power, coal, oil, hydro natural energy sources. The situation of energy in Vietnam was also mentioned in this paper. Vietnam has an abundant natural resources likes: coal, gas, hydro power etc, but from year 2013 to now Vietnam facing of electricity shortage and to solve the problem, Vietnam Government has chosen nuclear power energy to achieve energy balance between the rate of energy consumption and the ability to energy supply. Eight units will be built in Vietnam and in October 2014 Vietnamese officials have chosen Rosatom's AES-2006 design with VVER-1200/v-491 reactors for country's first nuclear power plant at Ninh Thuan and a second plant should follow based on a partnership with Japan. In this paper, the breakdown of NPP costs is considered. All the costs for building a NPP includes: the investment costs are the largest components (about 60%), fuel costs (15%), O and M costs (25%) and external costs are lower than 1% of the kWh costs. The situation for energy in Vietnam was mentioned with increase annually by 5.5 %, and now the shortage electricity is the big problem in power section. The purpose of this report is to give a general picture to consider the cost of nuclear power. It includes all the costs for building a nuclear power plant like total capital investment costs, production costs, external costs in which the capital investment costs is the largest component of the kWh cost. Nuclear energy Power was chosen to deal with situation of diminishing resources shortages.

  10. Improvement of automatic control systems of high-power turbines of PAO tubroatom for nuclear power plants

    Science.gov (United States)

    Shvetsov, V. L.; Babaev, I. N.

    2017-09-01

    The main technical solutions applied by PAO Turboatom used as the compensatory measures at the increase of the period of nonstop operation of nuclear power plants' (NPP) turbines with VVER-1000 type reactors up to 18 months are (1) replacing the standard hydraulic speed controller with an electronic one, (2) introduction of overclocking protection, (3) modernization of units of stop-control valves of high pressures, (4) installation of locking dampers on the receiver tubes of turbines of the first and second modification, and (5) improving the quality of repairs by reviewing the requirements for their implementation. The introduction of complex diagnostics of a control system on the basis of automatic treatment of results of registration of working parameters of the turbine is allocated as a separate prospective direction. Using an electronic controller of speed makes it possible to simplify the procedure of its inclusion in work at the failure of an electro-hydraulic system of control and vice versa. The regimes of maintaining the turbine rotor speed, steam pressure on the outlet of turbine, and the positions of main servomotors were introduced into the functions of the electronic controller. An electronic controller of speed includes its own electro-hydraulic transducer, turbine rotor speed sensor, and sensors of the position of main servomotors. Into the functions of electro- hydraulic control system and electronic speed controller, the function of overclocking protection, which determines the formation of commands for stopping the turbine at the exceeding of both the defined level of rotation speed and the defined combination of achieved rotation speed and angular acceleration of rotor, was introduced. To simplify the correction of forces acting on the control valve cups, the design of the cups was changed, and it has the profiled inserts. The solutions proposed were implemented on K-1100-60/1500-2M turbines of Rostov NPP. From the composition of control system

  11. Development of a new bench for puncturing of irradiated fuel rods in STAR hot laboratory

    Science.gov (United States)

    Petitprez, B.; Silvestre, P.; Valenza, P.; Boulore, A.; David, T.

    2018-01-01

    A new device for puncturing of irradiated fuel rods in commercial power plants has been designed by Fuel Research Department of CEA Cadarache in order to provide experimental data of high precision on fuel pins with various designs. It will replace the current set-up that has been used since 1998 in hot cell 2 of STAR facility with more than 200 rod puncturing experiments. Based on this consistent experimental feedback, the heavy-duty technique of rod perforation by clad punching has been preserved for the new bench. The method of double expansion of rod gases is also retained since it allows upgrading the confidence interval of volumetric results obtained from rod puncturing. Furthermore, many evolutions have been introduced in the new design in order to improve its reliability, to make the maintenance easier by remote handling and to reduce experimental uncertainties. Tightness components have been studied with Sealing Laboratory Maestral at Pierrelatte so as to make them able to work under mixed pressure conditions (from vacuum at 10-5 mbar up to pressure at 50 bars) and to lengthen their lifetime under permanent gamma irradiation in hot cell. Bench ergonomics has been optimized to make its operating by remote handling easier and to secure the critical phases of a puncturing experiment. A high pressure gas line equipped with high precision pressure sensors out of cell can be connected to the bench in cell for calibration purposes. Uncertainty analyses using Monte Carlo calculations have been performed in order to optimize capacity of the different volumes of the apparatus according to volumetric characteristics of the rod to be punctured. At last this device is composed of independent modules which allow puncturing fuel pins out of different geometries (PWR, BWR, VVER). After leak tests of the device and remote handling simulation in a mock-up cell, several punctures of calibrated specimens have been performed in 2016. The bench will be implemented soon in hot

  12. International Rivalry In The Energy Sector: The Eastern European Market Of Atomic Energy In Focus

    Directory of Open Access Journals (Sweden)

    Y. V. Borovsky

    2017-01-01

    Full Text Available In the post-bipolar world nuclear power has become one of the areas of competition and rivalry betweenRussiaand the West. The comprehensive analysis of theoretical publications allows us to consider international competition as an abstract, depoliticized contest of states and other international actors (including companies for some limited (mainly economic benefits. International rivalry is more a political process, necessarily involving some rival pairs of states (or groups of states that compete with each other not only to get some benefits, but to expand their territory or power. The competition and rivalry betweenRussiaand the West in the sphere of nuclear power are especially apparent in the Eastern European region where the American, European and Japanese corporations, with the support of the Western foreign ministries and EU institutions, try to achieve two main goals. The first goal is to win the contracts to build new power units, especially in tenders where Rosatom participates. The second goal is to become suppliers of nuclear fuel for multiple Russian- or Soviet-made VVER-type reactors, which are functioning or will be run in a number of countries in the region (Slovakia,CzechRepublic,Hungary,Bulgaria, andUkraine. Such activities can involve high risks. The West’s efforts to curb the dominant position of "Rosatom" inEastern Europeare formally associated with the need to create a "competitive market" of nuclear services in the region and to ensure the European energy security. It is also noteworthy that the expansion of Rosatom (and its predecessors to foreign markets, including Eastern Europe, is actively supported by the Russian state which in the second half of the 1990s – after a failed attempt of following in the footsteps of the West – joined in the rivalry, mostly imposed by the U.S. and their allies. As shown by the analysis,Russiaand the West, primarily theUnited States, are involved in the nuclear power sector to

  13. Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

    Science.gov (United States)

    Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

    2017-11-01

    The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification

  14. The gas turbine - modular helium reactor program for efficient disposition of weapons plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Kodochigov, N.G.; Kuzavkov, N.G.; Sukharev, Yu.P. [OKBM, Nizhny Novgorod (Russian Federation); Chudin, A.G. [MINATOM, Moscow (Russian Federation); Shenoy, A.S.; Simon, W.A. [General Atomics, San Diego, CA (United States)

    1998-09-01

    A large amount of weapons grade plutonium has been currently accumulated in the world. These stock-piles of accumulated plutonium are potentially hazardous because of the possibility of its unpermitted proliferation with subsequent manufacturing and use in nuclear weapons. From this point of view, the problem of the plutonium disposition is urgent. On the other hand, plutonium is an extremely valuable energy product, and should be used efficiently. The concept of burning WGPu in reactor power plants for electricity production is the official Russian position, and is considered as a long-term solution by existing power plants modification as well as with new reactor technologies development. Operating VVERs-1000 and BN-600 are some of the candidates to involve of WGPu in their fuel cycle, but the advantages of Gas Cooled High Temperature Reactors as GT-MHR allow to consider this type of reactor as a surplus WGPu consumer in the nearest future (2010). The inherent safety characteristics of the GT-MHR make it well suited to the mission of WGPu disposition. Because of the high burnup and no breeding of new plutonium, the GT-MHR consumes circa 90 % of the initially charged Pu-239. A single GT-MHR plant consisting of four reactor modules of 600 MWt power each can achieve this level of destruction for 50 tonnes of WGPu with concurrent electricity generation of circa 46 GW{center_dot}year over its design lifetime. In contrast, only 50 % of initial charged plutonium is consumed in LWR with electricity generation of circa 25 GW{center_dot}year. Discussion between General Atomics (GA) of United States and the Russian Ministry of Atomic Energy (MINATOM), in the summer of 1994, led to an agreement on a jointly funded design and development program for the GT-MHR, presented in a GA paper at this meeting. The program is initially focused on the burning of weapons plutonium that becomes available from dismantled nuclear weapons. The long term goal is to utilize the same design for

  15. ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    2017-06-19

    processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.

  16. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  17. Glasses for immobilization of low- and intermediate-level radioactive waste

    Science.gov (United States)

    Laverov, N. P.; Omel'yanenko, B. I.; Yudintsev, S. V.; Stefanovsky, S. V.; Nikonov, B. S.

    2013-03-01

    with high-power channel reactors (HPCR; equivalent Russian acronym, RBMK) and the Kalinin nuclear power plant with pressurized water reactors (PWR; equivalent Russian acronym VVER) after their 14-yr storage in the shallow-seated repository at the MosNPO Radon testing ground has confirmed the safety of repositories ensured by confinement properties of borosilicate matrix. The most efficient vitrification technology is based on cold crucible induction melting. If the content of a chemical element in waste exceeds its solubility in glass, a crystalline phase is formed in the course of vitrification, so that the glass ceramics become a matrix for such waste. Vitrified waste with high Fe; Na and Al; Na, Fe, and Al; Na and B is characterized. The composition of frit and its proportion to waste depends on waste composition. This procedure requires careful laboratory testing.

  18. Construction management of Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bohra, S.A. [Nuclear Power Corporation of India Limited, Vikram Sarabhai Bhavan, Anushaktinagar, Mumbai 400094 (India)]. E-mail: sabohra@npcil.co.in; Sharma, P.D. [Nuclear Power Corporation of India Limited, Vikram Sarabhai Bhavan, Anushaktinagar, Mumbai 400094 (India)

    2006-04-15

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  19. Coloss project; Le projet Coloss

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The COLOSS project was a shared-cost action, co-ordinated by IRSN within the Euratom Research Framework Programme 1998-2002. Started in February 2000, the project lasted three years. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H2 production, melt generation and the source term were studied, through a large number of experiments such as a) dissolution of fresh and high burn-up UO{sub 2} and MOX by molten Zircaloy, b) simultaneous dissolution of UO{sub 2} and ZrO{sub 2} by molten Zircaloy, c) oxidation of U-O-Zr mixtures by steam, d) degradation-oxidation of B{sub 4}C control rods. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics. Break-through were achieved on some issues. Nevertheless, more data are needed for consolidation of the modelling on burn-up effects on UO{sub 2} and MOX dissolution and on oxidation of U-O-Zr and B{sub 4}C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H2 production observed during the reflooding of degraded cores under severe accident conditions. Based on the experimental results obtained on the COLOSS topics, corresponding models were developed and were successfully implemented in several severe accident codes. Upgraded codes were then used for plant calculations to evaluate the consequences of new models on key severe accident sequences occurring in different plants designs involving B{sub 4}C control rods (EPR, BWR, VVER- 1000) as well as in the TMI-2 accident. The large series of plant calculations involved sensitivity studies and code benchmarks. Main severe accident codes in use in the EU for safety studies were used such as ICARE/CATHARE, SCDAP/RELAP5, ASTEC, MELCOR and MAAP4. This activity enabled: a) the assessment of codes to calculate core degradation, b) the

  20. Лексика песенных зачинов в белорусской свадьбе: семантика, структурные трансформации (на примере одной группы текстов

    Directory of Open Access Journals (Sweden)

    Галина [Galina] Кутырёва-Чубаля [Kutyriowa-Czubala

    2016-12-01

    : semantyka, transformacje strukturalne (na przykładzie jednej grupy tekstów W kontekście badań typologii dynamicznej oraz cech dialektalnych białoruskiej pieśni ludowej rozpatrujemy paradygmat tekstów dotyczących epizodu rozstawania się panny młodej z rodzicami. Są to teksty o tym samym podłożu leksykalno-poetyckim, które charakteryzuje wspólny typ incipitu w różnych jego wariantach. Wszystkie one symbolizują dramatyczny dla panny młodej moment „wtargnięcia” na jej podwórko „obcego” rodu – rodu narzeczonego. Metafora ta zawarta jest w formule incipitu. Warianty formuły zanotowano na znacznym obszarze Białorusi oraz na terenach przygranicznych (Łotwa, Polska. Transformacje strukturalne tego typu incipitów polegają na redukcji lub rozrastaniu (afiksacji składu sylabowego wersolinii. Wszystko to jest uwarunkowane semantyką pozawerbalną, wyrażoną przez rytmikę wokaliczną – z przewagą albo elementów deklamacyjności, albo dynamiki motoryczno-somatycznej. Charakterystyczne jest również poszerzenie wersu incipitu za pomocą dodawania do początkowego wyrazu synonimu rymowanego. Diachronia tekstów danego paradygmatu ujawnia się także przy zestawieniu tekstów tematycznie związanych z tymże epizodem wesela. Spora liczba tych tekstów przedstawia wersję skróconą – środkowy i końcowy fragmenty – tekstu pierwotnego danej grupy. Pieśń w wersji skróconej rozpoczyna się od punktu kulminacyjnego, z pominięciem metaforyczno-narracyjnej części wstępnej. Odpowiednio rolę incipitu pełni tu inna strofa. Teksty w wersji skróconej lokalizują się na peryferiach areału. Metodę dynamicznej typologii pieśni stosujemy do opisu tekstów różnych gatunków, o różnej tematyce, jednak należących ściśle do tego samego paradygmatu sylabo-ryt­micznego. Otwiera to nowe perspektywy w badaniach dialektów lingwo-muzycznych oraz zmian diachronicznych w języku pieśni ludowej.

  1. Fetysz „obiektywności”. „Nasza klasa” Tadeusza Słobodzianka

    Directory of Open Access Journals (Sweden)

    Tomasz Żukowski

    2016-01-01

     grają narzucone im role. Wątki te zostają jednak całkowicie pominięte przez krytykę. Nie są także w stanie przebić się przez lepiej rozpoznawalne społecznie struktury narracyjne pierwszego typu.

  2. Przestrzeń pożydowska

    Directory of Open Access Journals (Sweden)

    Konrad Matyjaszek

    2014-06-01

    Holokaustu oraz po roku 1945. Tego typu nieruchomości nazywa się obecnie „mieniem pożydowskim”. Autor analizuje dwa równolegle działające procesy adaptacji tej przestrzeni miejskiej. Pierwszy z nich zasadza się na koncepcji „przestrzeni żydowskiej”, przedstawionej w 1999 roku przez Dianę Pinto. „Przestrzeń żydowska”, pierwotnie zdefiniowana jako kulturalne i materialne miejsce spotkań europejskich Żydów oraz ludności pochodzenia nieżydowskiego, w polskim kontekście nie wymaga realnej obecności Żydów. Co za tym idzie, praktyki społeczne wynikające z tej interpretacji idei „przestrzeni żydowskiej” daleko odbiegają od pierwotnego zamysłu Pinto. Drugi proces opisany w artykule dotyczy fizycznej realizacji tak zdefiniowanego „miejsca spotkań”. Analizowany jest on na przykładzie Dotleniacza – instalacji miejskiej z 2007 roku autorstwa Joanny Rajkowskiej. Praca Rajkowskiej była jedną z pierwszych prób stworzenia fizycznego miejsca spotkań w Polsce. Pomimo altruistycznych założeń, które legły u podstaw projektu, nie mógł on w pełni wyrwać się z okowów dyskursu na temat „dialogu” ekskluzywistycznego. Tak więc kulturowa interpretacja Dotleniacza została zawężona do znaczeń, których owocem będzie raczej wykluczenie, a nie możliwość uczestnictwa.

  3. Próba analizy zjawiska konwergencji na przykładzie działań grupy ITI

    Directory of Open Access Journals (Sweden)

    Grzegorz Walczak

    2011-05-01

    Full Text Available Konwergencja mediów może zachodzić na wielu płaszczyznach: technologicznej, kulturowej, społecznej, czy też przemysłowej. Proces ten opisuje zmiany występujące zarówno w sferze produkcji, jak i konsumpcji przekazów medialnych. Wspólnym mianownikiem konwergencji – jak zauważają medioznawcy – jest przekaz cyfrowy. Dzięki digitalizacji przekazu analogowego na cyfrowy, różne urządzenia upodabniają się do siebie, świadczą te same usługi i dostarczają tych samych treści swoim odbiorcom. Koncentracja mediów w obrębie jednej grupy kapitałowej pozwala na praktyczne wykorzystanie tego procesu. Dysponując wieloma środkami przekazu można tworzyć różnego rodzaju powiązania między udostępnianą zawartością, a tym samym wypracowywać zupełnie nowy model biznesowy w zakresie produkcji, rozpowszechniania oraz promocji własnych zasobów medialnych. W ITI doskonale zdawano sobie sprawę z korzyści wynikających z właściwego wykorzystania procesu konwergencji. Z tego też powodu systematycznie wkraczano na zupełnie nowe pola podejmowanej działalności. Funkcjonowanie koncernu na rynku medialnym opierało się na kilku zasadniczych filarach: ogólnodostępnej telewizji TVN, kanałach tematycznych, portalu internetowym Onet oraz powiązanych z nim witrynach, a także na platformie cyfrowej „n”. Wszystkie te podmioty wchodziły ze sobą w mniej lub bardziej złożone interakcje. Zawartość produkowana na potrzeby jednego medium „przenikała” do drugiego i wykorzystując jego specyfikę była w pewnym stopniu rozbudowywana. Chodziło o to, aby ten sam produkt medialny można było sprzedawać na wiele zupełnie różnych sposobów. W ten sposób starano się przedłużać życie marki, budować emocjonalne przywiązanie odbiorców do danej zawartości, stosować narrację transmedialną, czy też kontrolować przepływ widowni. Tego typu działalność miała swoje ekonomiczne uzasadnienie, bowiem pozwala

  4. VZTAH MEZI VĚDOMOSTMI O PROBLEMATICE POHYBOVÉ AKTIVITY A REALIZOVANOU POHYBOVOU AKTIVITOU U STŘEDOŠKOLSKÝCH STUDENTŮ [RELATIONSHIP BETWEEN KNOWLEDGE ABOUT PHYSICAL ACTIVITY AND PERFORMED PHYSICAL ACTIVITY IN HIGH SCHOOL STUDENTS

    Directory of Open Access Journals (Sweden)

    Jana Vašíčková

    2009-12-01

    Full Text Available VÝCHODISKA: Školní vzdělávací programy a změny výuky ve školách nabízejí prostor pro uchopení tématu zdravého životního stylu v životě člověka, a to jak po stránce zdravotní, nutriční, tak i pohybové. CÍLE: Cílem práce bylo zjistit, jaká je úroveň teoretických vědomostí o problematice zdraví a pohybové aktivity (PA, zda se prokáže rozdíl mezi chlapci a dívkami a do jaké míry existuje spojitost mezi vědomostmi a uskutečněným množstvím PA v rámci současných kurikul u studentů prvního ročníku středních škol. METODIKA: Výzkumného šetření se zúčastnilo 33 chlapců a 42 dívek prvního ročníku ze tří gymnázií a jedné střední odborné školy. Tito studenti v úvodu šetření vyplnili vědomostní test (VT k problematice zdraví a PA a dotazník ANEWS (Abbreviated Neighborhood Environment Walkability Scale. Poté po celý měsíc monitorovali krokoměrem Yamax Digiwalker SW-700 množství denních kroků, které zapisovali do brožury. Na závěr, po týdenní pauze od monitorování, studenti opět vyplnili vědomostní test a dotazník IPAQ (International Physical Activity Questionnaire – dlouhou verzi. Při podrobné analýze vztahu mezi vědomostmi a PA jsme se zaměřili také na PA různé intenzity a typu. VÝSLEDKY: Zjistili jsme, že dívky mají větší znalosti o problematice PA než chlapci ve všech sledovaných dimenzích, a to jak v pre-testu (18,26 vs. 16,55 bodů, tak i v post-testu (18,57 vs. 16,76 bodů. Korelační koeficient u chlapců mezi celkovou PA a VT byl rP = - 0,505 (p[BACKGROUND: School education programs and changes in teaching at schools offer the possibility to emphasize the topic of healthy lifestyle in human life with the impact on health, nutrition, and physical activity. OBJECTIVE: The main aim of this study was to find out the level of theoretical knowledge about health and physical activity (PA, the difference between genders and whether

  5. Modelling of Underground Coal Gasification Process Using CFD Methods / Modelowanie Procesu Podziemnego Zgazowania Węgla Kamiennego Z Zastosowaniem Metod CFD

    Science.gov (United States)

    Wachowicz, Jan; Łączny, Jacek Marian; Iwaszenko, Sebastian; Janoszek, Tomasz; Cempa-Balewicz, Magdalena

    2015-09-01

    , under the specific conditions of the georeactor operations within the time interval of 100 hours and 305 hours. The results of the numerical solution have been compared with the results of experimental results under in-situ conditions. Zaprezentowano wyniki badań modelowych polegających na numerycznej symulacji procesu podziemnego zgazowania węgla. Dla potrzeb realizowanej pracy dokonano wyboru oprogramowania wykorzystywanego do symulacji procesu podziemnego zgazowania węgla. Na podstawie przeglądu literatury zdecydowano, że oprogramowaniem, za pomocą, którego będą realizowane badania modelowe, będzie oprogramowanie informatyczne ANSYS-Fluent. Za jego pomocą przeprowadzano obliczenia numeryczne z zamiarem zidentyfikowania rozkładu zmian stężenia składników gazu procesowego w funkcji czasu trwania procesu zgazowania węgla. Przeprowadzone obliczenia miały charakter predykcji. W oparciu o dane konstrukcyjne georeaktora stosowanego podczas badań na KD Barbara oraz KWK Wieczorek, opracowano model geometryczny oraz wykonano jego dyskretyzację poprzez wygenerowanie odpowiedniej siatki numerycznej w oparciu, o którą wykonywane są obliczenia. Dane dotyczące sposobu zasilania georeaktora oraz parametrów utrzymywanych podczas procesu wykorzystano do definiowania modelu numerycznego. Część danych została uzupełniona w oparciu o źródła literaturowe. Głównym przyjętym założeniem było oparcie symulacji pracy georeaktora o modele opisujące reaktywny przepływ płynu. Składniki gazu procesowego oraz czynnik zgazowujący przemieszczają się wzdłuż kanału zgazowującego symulując zjawiska fizykochemiczne związane z transportem masy i energii oraz zachodzące reakcje chemiczne (wraz z efektem energetycznym). Chemizm procesu zgazowania oparto o równanie kinetyczne, które determinuje przebieg danego typu równania chemicznego zgazowania węgla. W ramach modelu opisano też interakcję gazu z otaczającą warstwą węgla. Opis ten dotyczy

  6. The ​Fest-noz: a Way to Live Breton Culture

    Directory of Open Access Journals (Sweden)

    Nicole Dołowy-Rybińska

    2015-06-01

    Full Text Available The Fest-noz: a Way to Live Breton Culture The history of the Breton language and culture of the XX and XXI century is narrated through the fest-noz phenomenon story. Fest-noz (‘night festival’ is a meeting where people dance in groups to live folk music accompaniment. Traditionally these festivals were organized in the small region of Central Brittany and were connected to important community occasions. This tradition was slowly disappearing in the 20’s and 30’s of the XXth century to die out after WW II when the Breton culture was depreciated and connected with the negative identity of the Bretons. Fest-noz was recreated and in 50’s and has become an invented tradition. In the 60’s with the social and cultural movements leading to the revalorization of the minority fest-noz became the symbol of a Breton ethnic revival. Today it is one of the most significant marks of Breton identification. Every year there are hundreds fest-noz organized all around Brittany, from small local celebrations to huge musical events. Participation in fest-noz is one of the ways of conscious creation of the Breton cultural identity.   Fest-noz: sposób na życie kulturą bretońską Fest-noz („nocna zabawa” to spotkanie, podczas którego ludzie tańczą wspólnie do muzyki bretońskiej. Tradycyjnie zabawy takie odbywały się na niewielkim obszarze Bretanii środkowej i związane były z ważnymi wydarzeniami wspólnot wiejskich (związanych z uprawą ziemi, ale i obchodami świąt religijnych czy prywatnych. Brała w nich udział cała społeczność okolicy, tańcząc w łańcuchach do improwizowanych pieśni typu kan ha discan („zawołanie i odpowiedź”. Tradycja ta powoli zamierała w latach 30. i 40. XX wieku, zaś po II wojnie światowej fest-noz nie odbywało się wcale. W latach 60. XX wieku fest-noz zostało w sposób celowy zrewitalizowane, by stać się symbolem bretońskiego odrodzenia etnicznego przełomu lat 60. i 70., buntu m

  7. Биология vs. социология и напредъкът на националното тяло (Един теоретичен дебат в българското общество от междувоенния период

    Directory of Open Access Journals (Sweden)

    Нина [Nina] Димитрова [Dimitrova

    2017-10-01

    żliwościom socjologii biologicznej, nowego trendu w nauce, w odniesieniu do potencjalnych korzyści dla postępu narodu – kwestii priorytetowej w okresie międzywojennym. Autor przedstawia argumenty bułgarskich filozofów, socjologów, psychologów, psychiatrów oraz innych myślicieli na polu humanistyki, którzy komentują osiągnięcia biologii i medycyny i ich zastosowanie do „polepszenia” człowieka i uzdrowienia życia społecznego na drodze różnego typu eugeniki negatywnej. Uwagę skupiono na sporach teoretycznych między dwoma podejściami do społeczeństwa; ich poważne konsekwencje są komentowane przede wszystkim w kontekście samej debaty. Dyskusja jest ujęta również jako echo ówczesnych trendów europejskich, dotyczących nowej roli biologii. Dwie szkoły filozoficzne w Bułgarii – remkeanizm i marksizm – popierały „socjologię autonomiczną”, wedle której społeczeństwo ludzkie jest radykalnie różne od sfery natury, tak więc czynnik biologiczny nie jest dla zmian społecznych istotny. Zgodnie z przeciwnym punktem widzenia, częściej spotykanym, ważniejsze niż wpływ środowiska i wychowania jest dziedziczenie, toteż eugenika – oparta na genetyce nauka o polepszeniu człowieka i ludzkości – daje wiele możliwości. W artykule ukazano różnorodne stanowiska, pojawiające się w wyniku zmiany intelektualnego paradygmatu „biologizowania” nauk społecznych, jak również autentyczne postawy leżące u ich podstaw. Autor analizuje ślady debaty także w dzisiejszych czasach.

  8. Elektivní chirurgické zvyšování výkonu u atletů: Měli bychom se mu bránit? Elective performance enhancement surgery for athletes: Should it be resisted?

    Directory of Open Access Journals (Sweden)

    Mark Hamilton

    2006-02-01

    Full Text Available Následující článek popisuje některé chirurgické zákroky, které se používají za účelem zvyšování sportovního výkonu, a přináší diskusi o některých možných zákrocích, které by mohly být pro tyto účely používány v budoucnosti. U mnoha atletů ve sportovních odvětvích, ve kterých je zásadním předpokladem úspěchu zrak, se provádějí elektivní oční chirurgické zákroky. Vyvstává tím etická otázka, zda je morálně přijatelné provádět zákroky za účelem rozvoje schopností, které přesahují normu, jako je například přesnost zraku 20/10. Nezbytná jsou kritéria pro stanovení morálnosti těchto zákroků, zvláště pak pokud je chirurgický zákrok volitelný, který není motivován terapeutickými potřebami, nýbrž snahou zvýšit výkon a vytvořit schopnosti přesahující normu. Jednou věcí je podstoupit korektivní chirurgický zákrok pro vlastní pohodlí nebo za účelem zpomalení postupu zhoršování, naprosto jinou věcí je pak tento zákrok za účelem dosažení dokonalého zraku jako je 20/10. Přijmeme-li tento spíše neškodný oční chirurgický zákrok a uvážíme-li pokroky v oblasti minimálně invazivních chirurgických metod, přísná opatření proti používání steroidů a stále rostoucí finanční impulsy pro sportovce, je pouze otázkou času, kdy už půjde o to, aby sportovci podstupovali elektivní chirurgické zákroky za tím účelem, aby se stali většími, silnějšími nebo rychlejšími. Elektivní chirurgie se v oblasti zvyšování výkonu rozšíří. Vyvstává tím množství morálních problémů, a proto musíme seriózně začít hledat kritéria, pomocí nichž bude možno morálnost tohoto typu elektivní chirurgie posuzovat, a povzbuzovat sportovní instituce, aby proaktivně vytvářely příslušné strategie. Konečně je zde diskuse o tom, zda lze řešení hledat v transhumanismu, v pragmatickém přístupu nebo v

  9. Profil ekspresji genów w gwiaździakach włosowatokomórkowych wieku dziecięcego w odniesieniu do lokalizacji, obrazu radiologiczno‑morfologicznego i przebiegu klinicznego choroby

    Directory of Open Access Journals (Sweden)

    dr n. med. Magdalena Zakrzewska

    2011-10-01

    martwicy centralnej i przebiegu klinicznego choroby (5 przypadków z cechami klinicznymi progresji choroby po resekcji subtotalnej, 2 przypadki rozwijające się w przebiegu neurofibromatozy typu 1.. Po normalizacji wyników przy użyciu algorytmu GC­‑RMA przeprowadzono analizy bioinformatyczne wykorzystujące przede wszystkim pakiet Bioconductor. Wyselekcjonowano 862 geny różnicujące gwiaździaki włosowatokomórkowe pod względem umiejscowienia anatomicznego i wykazano obecność istotnej zależności statystycznej pomiędzy profilem ekspresji genów w odniesieniu do lokalizacji zmiany (p=0,001. Na podstawie uzyskanych wyników dokonano wyboru genów będą­ cych markerami molekularnymi dla nowotworów rozwijających się w poszczególnych lokalizacjach (IRX2, PAX3, CXCL14, LHX2, SIX6, CNTN1, SIX1. Nie wykazano możliwości zróżnicowania badanej grupy w zależności od obrazu radiologiczno­‑morfologicznego. Geny najlepiej różnicujące badaną grupę cechowały się małą amplitudą zmian i brakiem znamienności statystycznej (p=0,88. Podobnie progresja choroby nie była związana z profilem ekspresji genów (p=0,83. Walidację uzyskanych wyników przeprowadzono w oparciu o QRT­‑PCR. Przeprowadzone analizy pozwoliły stwierdzić, że gwiaździaki włosowatokomórkowe w zależności od lokalizacji anatomicznej posiadają charakterystyczny profil ekspresji genów, sugerujący ich różne pochodzenie. Z kolei obraz radiologiczno­‑morfologiczny oraz przebieg kliniczny choroby nie mają związku z całkowitym profilem ekspresji genów.

  10. Properties of Waste from Coal Gasification in Entrained Flow Reactors in the Aspect of Their Use in Mining Technology / Właściwości odpadów ze zgazowania węgla w reaktorach dyspersyjnych w aspekcie ich wykorzystania w technologiach górniczych

    Science.gov (United States)

    Pomykała, Radosław

    2013-06-01

    Most of the coal gasification plants based of one of the three main types of reactors: fixed bed, fluidized bed or entrained flow. In recent years, the last ones, which works as "slagging" reactors (due to the form of generated waste), are very popular among commercial installations. The article discusses the characteristics of the waste from coal gasification in entrained flow reactors, obtained from three foreign installations. The studies was conducted in terms of the possibilities of use these wastes in mining technologies, characteristic for Polish underground coal mines. The results were compared with the requirements of Polish Standards for the materials used in hydraulic backfill as well as suspension technology: solidification backfill and mixtures for gob caulking. Większość przemysłowych instalacji zgazowania węgla pracuje w oparciu o jeden z trzech głównych typów reaktorów: ze złożem stałym, dyspersyjny lub fluidalny. W zależności od rodzaju reaktora oraz szczegółowych rozwiązań instalacji, powstające uboczne produkty zgazowania mogą mieć różną postać. Zależy ona w dużej mierze od stosunku temperatury pracy reaktora do temperatury topnienia części mineralnych zawartych w paliwie, czyli do temperatury mięknienia i topnienia popiołu. W ostatnich latach bardzo dużą popularność wśród instalacji komercyjnych zdobywają reaktory dyspersyjne "żużlujące". W takich instalacjach żużel jest wychwytywany i studzony po wypłynięciu z reaktora. W niektórych przypadkach oprócz żużla powstaje jeszcze popiół lotny, wychwytywany w systemach odprowadzania spalin. Może być on pozyskiwany oddzielnie lub też zawracany do komory reaktora, gdzie ulega stopieniu. Wszystkie z analizowanych odpadów - trzy żużle oraz popiół pochodzą właśnie z tego typu instalacji. Tylko z jednej z nich pozyskano zarówno żużel jak i popiół, z pozostałych dwóch jedynie żużel. Odpady te powstały, jako uboczny produkt zgazowania w

  11. Úvodník 2010/V/1

    Directory of Open Access Journals (Sweden)

    Jana Dlouhá

    2010-05-01

    čními. Ve svém krátkém článku poukazuje na paralely mezi matematikou a environmentální výchovou ‑ nachází je v pěstování kultivovaného vztahu ke skutečnosti. Pokora, nenásilí, odříkání se přemíry materiálního blahobytu, to jsou ale hodnoty, které v naší kultuře těžko nazvat alternativními. A to nás zase může svést k myšlenkám o základech hodnot ekonomických……protože totiž odměna a úspěch se až nápadně často nedostaví tam, kde se po nich člověk pídí v první řadě. Což vysvítá i mezi řádky rozhovoru s Lubomírem Nátrem, kterého jsme oslovili jako prvního z „nestorů“ (nejen environmentálního (a nejen vzdělávání. Budeme-li chtít, dočteme se zde, že věda může být především okouzlením a pobytem ve společenství stejně zaujatých, věřících i pochybujících ohledávačů jsoucího; těch, pro které pochopit a sdělit, obstát v kritickém pohledu, znamená víc, než ovládat, vykládat, vlastnit, manipulovat. Předpokladem úspěchu je (i zde pokora, poctivost, a schopnost (sebekritiky; a my můžeme opět spekulovat o tom, že právě tyto hodnoty jsou prvním a legitimním cílem lidského snažení, který ale podmiňuje zdar při dosahování cílů jiných, sekundárních.Po této inspirativní duchovní rozcvičce se už ale začteme do odborných textů tohoto čísla. Jan Činčera, Petr Gilar a Júlia Sokolovičová analyzují vzdělávací kurzy pro koordinátory environmentální výchovy; ukazují, že vlastní hodnocení studentů nemusí na rozdíl od expertního pohledu odhalit nedostatky v naplňování vzdělávacích cílů tohoto typu studia ‑ ty ovšem mohou souviset s deficity v pregraduální přípravě učitelů. Ve škole zůstaneme: Milena Pouchová se zabývá projektovým vyučováním z hlediska teoretického a představuje také výzkum provedený na reprezentativním vzorku základních škol. Ukazuje rozpor mezi teorií a praxí tohoto

  12. New gravity control in Poland - needs, the concept and the design

    Science.gov (United States)

    Krynski, Jan; Olszak, Tomasz; Barlik, Marcin; Dykowski, Przemyslaw

    2013-06-01

    transformation to a new system (as 2nd order network) as well as a definition of gravity system as "zero-tide" system. Seasonal variability of gravity has been discussed indicating that the effects of environmental changes when establishing modern gravity control with absolute gravity survey cannot be totally neglected. Założona w Polsce w ostatniej dekadzie XX wieku zgodnie z obowiązującymi standardami międzynarodowymi Podstawowa Osnowa Grawimetryczna Kraju (POGK), składająca się z około 350 punktów, została oparta na 12 absolutnych punktach grawimetrycznych, na których przyspieszenie siły ciężkości wyznaczono przy użyciu czterech różnych typów grawimetrów absolutnych. Względne pomiary grawimetryczne na punktach tej osnowy, z jednoczesnym dowiązaniem jej do przyspieszenia siły ciężkości na 12 absolutnych punktach grawimetrycznych, wykonały różne grupy pomiarowe przy wykorzystaniu grawimetrów LaCoste&Romberg (LCR). Konstrukcja powstałej sieci grawimetrycznej, w szczególności ograniczona liczba nierównomiernie rozłożonych punktów absolutnych na terenie kraju, na których w dodatku przyspieszenie siły ciężkości wyznaczono różnymi instrumentami w różnych epokach, spowodowały wystąpienie błędów systematycznych w wartościach g na punktach POGK. W niniejszej pracy, przy wykorzystaniu pomiarów grawimetrycznych wykonanych w latach 2007-2008 dokonano oceny tych błędów oraz przeprowadzono dyskusję ich możliwych źródeł. Rozwój technologii absolutnych pomiarów grawimetrycznych, w szczególności instrumentów przeznaczonych do precyzyjnych absolutnych pomiarów grawimetrycznych w warunkach polowych, stwarza możliwość założenia nowego typu osnowy grawimetrycznej, składającej się ze stacji, na których przyspieszenie siły ciężkości jest pomierzone grawimetrami absolutnymi. Nowa osnowa grawimetryczna Polski, która będzie zakładana w latach 2012-2014, będzie się składała z 28 punktów fundamentalnych (mierzonych

  13. Influence of forest management on the changes of organic soil properties in border part of Kragle Mokradlo Peatland (Stolowe Mountains National Park, Poland)

    Science.gov (United States)

    Bogacz, A.; Roszkowicz, M.

    2009-04-01

    . CONCLUSIONS Shallow organic soils occupy the ombrotrophic sites of a border part of Kragle Mokradlo Peatland. The variety of organic soil throphism in the object resulted from the placement on the base sandstone, partial mixing of soil horizons as well as from muddy and fluvial processes. Peat horizons present in the studied profiles were generally classified as hemic and sapric, sometimes as fibric. Soil horizons exhibited differed thickness and ash content. Forest management strongly changed the properties of organic soil. REFERENCES Bogacz, A. (2000). Physical properties of organic soil in Stolowe Mountains National Park (Poland). Suo 51,3; pp.105-113. Gee, G.W. and Bauder, J.W. (1986). Particle-size analysis. In: Klute, A.(ed.) Methods of Soil Analysis Part I. Agronomy series No. 9. Am. Soc. Agronomy Soil Sci. Am, Inc., Publ., Madison, WI.pp. 383-411. Horawski, M. (1987). Torfoznawstwo dla meliorantow. Pojecia podstawowe.[Peat science for melioration. Basic definitions.]. Wydawnictwo Akademii Rolniczej w Krakowie. pp.37-39.[In Polish]. Lubliner - Mianowska, K. (1951). Wskazowki do badania torfu. Metody geobotaniczne, polowe i laboratoryjne [Hints to peat research. In: Geobotanical, field and laboratory methods] Państwowe Wydawnictwo Techniczne, Katowice.pp.58-60. [In Polish]. Lynn, W.C., Mc Kinzie, W.E., Grossman, R.B. (1974). Field Laboratory Test for characterization of Histosols. In: Histosols, their characteristics, classification and use. pp. 11-20. Oznaczanie gatunku, rodzaju i typu torfu. (1977). [Peat and peat varies. Determination of classes, sort and types of peat]. Polish standard PN-76/G-02501, [Polish Normalization Commitee]. pp.1-11.[In Polish]. Word Reference Base for Soil Resources. 1998. Word Soil Resources Report, 84. FAO-ISRIC-ISSS, Rome, pp.1-88. Zawadzki, S. (1970). Relationship between the content of organic matter and physical properties of hydrogenic soils. Polish Journal of Soil Science Vol.III, 1; pp.3-9.

  14. Vztah mezi úspěšností dětí v předplavecké výchově, jejich temperamentovými charakteristikami a stimulací k pohybovým aktivitám Relationship between children's successfulness in pre-swimming education, their temperament characteristics and stimulation to physical activities

    Directory of Open Access Journals (Sweden)

    Ludmila Miklánková

    2006-02-01

    skupiny U1 také v typu temperamentu (p < 0.05. U sledovaného souboru byla prokázána statisticky významná závislost mezi úspěšností dítěte v předplavecké výchově a typem temperamentu (p < 0.05. The main aim of this inquiry was to review and evaluate possibilities of relations of preschool children's successfulness in a pre-swimming course to some outer and inner factors. We concentrated on a level of stimulation to physical activities from family and school and on temperament characteristics of monitored children. Our sample consisted of 83 children (non-swimmers attending kindergartens. Complete results were obtained from 58 of them (30 girls and 28 boys. The mean age of the sample was 5.87 years. A degree of successfulness of a child in the pre-swimming education course, concentrated on teaching swimming fundamentals, was evaluated by a standardized 5 item set of tests (Řehoř, 1969. The sample was divided into three groups according to a total achieved test score as follows: U1 – very successful, U2 – successful, and U3 – unsuccessful. Temperament characteristics (temperament type and character dimensions were assessed by the Eysenck questionnaire (Eysenck & Eysenck, 1994. The ESPA questionnaire (Renson & Vanreusel, 1990 was used to evaluate environmental stimulation to physical activities. A relationship between children's successfulness in pre-swimming education, their temperament type and character dimensions was evaluated by analysis of variance or the Kruskal-Walis test, respectively. The Mann-Whitney test was used to evaluate associations between stimulation to physical activities and successfulness in the pre-swimming education course. When comparing stimulation to physical activities between groups with different levels of successfulness in pre-swimming education (between groups U1 and U3 or U2 and U3, respectively, a significant difference was found (p < 0.01 in one of the stimuli of social participation. A comparison between the U1, U2

  15. Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Dassen, Lars van; Delalic, Zlatan; Ekblad, Christer; Keyser, Peter; Turner, Roland; Rosengaard, Ulf; German, Olga; Grapengiesser, Sten; Andersson, Sarmite; Sandberg, Viviana; Olsson, Kjell; Stenberg, Tor

    2009-10-15

    The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral assistance to Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in various projects financed by the European Union. The purpose of this project-oriented report is to provide the Swedish Government and other funding agencies as well as other interested audiences in Sweden and abroad with an encompassing understanding of our work and in particular the work performed during 2008. the activities are divided into four subfields: Nuclear waste management; Reactor safety; Radiation safety and emergency preparedness; and, Nuclear non-proliferation. SSM implements projects in the field of spent nuclear fuel and radioactive waste management in Russia. The problems in this field also exist in other countries, yet the concentration of nuclear and radioactive materials are nowhere higher than in north-west Russia. And given the fact that most of these materials stem from the Cold War era and remain stored under conditions that vary from 'possibly acceptable' to 'wildly appalling' it is obvious that Sweden's first priority in the field of managing nuclear spent fuel and radioactive waste lies in this part of Russia. The prioritisation and selection of projects in reactor safety are established following thorough discussions with the partners in Russia and Ukraine. For specific guidance on safety and recommended safety improvements at RBMK and VVER reactors, SSM relies on analyses and handbooks established by the IAEA in the 1990s. In 2008, there were 16 projects in reactor safety. SSM implements a large number of projects in the field of radiation protection and emergency preparedness. The activities are at a first glance at some distance from the activities covered and

  16. Decommissioning in western Europe; Kaernkraftsavveckling i Vaesteuropa

    Energy Technology Data Exchange (ETDEWEB)

    Lundqvist, K. [Castor arbetslivskonsulter AB, Stockholm (Sweden)

    1999-12-01

    and waterproof conditions for a longer period of time (sometimes hundred years or more), prior to final demolition. Among the reasons for deferring the dismantling are lack of waste repositories and decreasing dose-rates for the workers. Of Europe's 218 commercial reactors in operation, the majority, 151, are located i the Western part. The biggest producers are France, United Kingdom and Germany, with 58, 35 and 20 reactors respectively. Until now mostly research- and pilot reactors have been shut-down. There are yet few experiences from decommissioning of large-scale commercial reactors. The following commercial reactors are undergoing decommissioning. (There are also a great amount of nuclear facilities of other types being decommissioned.) The three gas-cooled twin reactor plants of Berkeley, Trawsfynydd and Hunterston in UK. In Germany Gundremmingen, Lingen, Kahl and Wuergassen are being decommissioned. All of them are located in the Western part of the country. The biggest project is however the dismantling of the gigantic Greifswaldfacility situated on the coast of the Baltic see in former Eastern Germany. The plant has eight Russian built reactors of VVER-type. Like the rest of the former GDR-plants Greifswald was shutdown after the reunification in 1990. The strategy chosen is immediate dismantling. France is decommissioning seven reactors (Chooz A1, Chinon A1, A2, A3, St Laurent A1, A2 and Bugey 1.) The oldest, Chinon A1, closed down in 1973 and the youngest, Bugey 1, in 1994. Italy closed down all NPPs (altogether four) in 1987 after a referendum. The first reactor of the Netherlands was shutdown in 1997 mainly for economical reasons. The development of a free European electricity market will make it less profitable to run certain facilities. Vandelos 1 in Spain is undergoing decommissioning after a fire in the turbines in 1989. IAEA, OECD/NEA and EU are co-operating in the field of decommissioning. Much work is spent on harmonizing rules and preparing

  17. Působení kognitivně-behaviorální psychoterapie na tělesné složení a konstituci Effects of cognitive behavioral psychotherapy on body composition and constitution

    Directory of Open Access Journals (Sweden)

    Miroslav Kopecký

    2008-01-01

    žší průměrná hodnota se opět nacházela u skupiny nejmladších žen (83,6, nejvyšší (88,6 u skupiny nejstarších žen. Na základě hodnocení množství podkožního tuku dle metodiky Matiegky 57,5 % žen disponovalo více než 30 % tuku, z toho v kategorii nad 40 % podkožního tuku se nacházelo 28,75 % souboru. Hodnoty indexů centrality (1,3–1,5, které dokumentují rozložení podkožního tuku v jednotlivých oblastech těla, korespondují s vyššími hodnotami WHR a potvrdily především uložení tuku na trupu vzhledem ke končetinám. Byl tedy potvrzen výskyt abdominálního typu obezity. Na základě segmentální analýzy realizované metodou bioelektrické impedance bylo nejvíce tuku determinováno na dolních končetinách. Intervence prostřednictvím pohybové aktivity a změny výživových zvyklostí má především individuální dopad. Na základě hodnocení průměrných hodnot obvodových parametrů došlo ke snížení především obvodu pasu, břicha a boků, případně ke snížení obvodových parametrů na dolních končetinách (obvody gluteálního a středního stehna. Snížily se také průměrné hodnoty tělesné hmotnosti, BMI a WHR a zastoupení množství tukové složky v absolutních i relativních hodnotách. Ve všech kategoriích jsme zaznamenali signifikantní snížení endomorfie. Velmi významnou vlastností je silná vůle, která je nedílnou součástí intervence a kterou ne každá žena disponuje. Takže vždy v kurzech nacházíme i malé procento žen, které nesníží ani hmotnost, ani nedojde ke snížení obvodových parametrů a množství podkožního tuku zůstane stejné jako před nástupem na terapii. Obesity is a chronic disease of modern times that is not just a cosmetic problem but a bio-social-psychological problem especially, which means that obese individuals have, apart from their medical problems, also social and psychological problems. They often suffer from depression, low self

  18. Assessment of postural stability in patients with a transtibial amputation with various times of prosthesis use [Hodnocení posturální stability pacientů s transtibiální amputací s různou dobou používání protézy

    Directory of Open Access Journals (Sweden)

    Dagmar Kozáková

    2009-09-01

    určení základních parametrů posturální stability byly použity dvě silové plošiny Kistler (typ 9286AA. Stabilita byla testována po dobu 30 s ve 4 modifikacích stoje (přirozený bipedální stoj, bipedální stoj s úzkou bázi, přirozený bipedální stoj se zavřenýma očima a stoj na molitanu. Pro určení vlivu doby používání protézy na úroveň posturální stability jsme použili korelační analýzu. Rozdíl mezi jednotlivými modifikacemi stoje byl hodnocen analýzou rozptylu pro opakovaná měření a LSD post hoc testem. VÝSLEDKY: Zatížení na zdravé končetině je u osob s transtibiální amputací ve všech typech stoje větší než na postižené končetině (rozdíl 17,8 až 21,8 % v závislosti na typu stoje. To platí také pro velikost výchylky COP v mediolaterálním směru a pro rychlost pohybu COP v anteroposteriorním a v mediolaterálním směru (p < 0,01, p < 0,05. Parametry charakterizující pohyb COP korelují (p < 0,01 na postižené končetině s rychlostí pohybu COP (s výjimkou stoje na molitanu, p < 0,01. Na zdravé končetině platí tato závislost pouze pro pohyb COP v mediolaterálním směru. Mezi jednotlivými typy stoje (s výjimkou přirozeného stoje jsme nenalezli významné rozdíly v rozsahu a v rychlosti pohybu COP. ZÁVĚRY: S rostoucí dobou, která uplyne mezi amputací a vybavením protetickou pomůckou, dochází k nárůstu asymetrie v zatížení amputované a zdravé končetiny, rozsah pohybu COP a jeho rychlost se zvětšují. Pro zmenšení pravděpodobnosti přetěžování zdravé končetiny v bipedálním stoji je nutné využít všechny možnosti pro zkrácení doby při vybavení protézou.

  19. Úvodník 2010/V/3

    Directory of Open Access Journals (Sweden)

    Jana Dlouhá

    2010-12-01

    co by to mělo být? Zlatá rybko, pověz, co bys uměla zařídit: zahraniční verzi časopisu, mezinárodní redakční radu, uznání ve světových databázích SCOPUS či WOS? Ne, rybko, vrať se do moře, my budeme skromní. Tohle všecko dokážeme sami, pokud  … budeme prostě existovat. Nechť se tak stane! Přejeme si tedy, abyste nám přáli vy i okolnosti; navázat snad máme na co ... [viz 5]A nyní - co toto číslo přináší. Autoři Jan Činčera, Kateřina Kohoutová a Julie Sokolovičová předkládají výsledky první fáze evaluačního výzkumu realizovaného mezi účastníky specializačního studia pro koordinátory environmentální výchovy – ukazují, jak studenti reflektují probíhající studium. Svatava Janoušková, Tomáš Hák, Jan Maršák a Lenka Pachmanová podávají přehled indikátorů vzdělávání pro udržitelný rozvoj na mezinárodní úrovni a navrhují, jak doplnit existující indikátorové sady. Text Vzdělávání metodou e-learningu na vysoké škole… rekapituluje možnosti, které tento typ vzdělávání nabízí pro podporu mezinárodní spolupráce a interdisciplinární komunikace; je navíc propojen s metodikami zpracovávajícími možné praktické postupy. Poslední příspěvek recenzované rubriky je už rázu zcela metodického: týká se způsobů evaluace programů environmentální výchovy, které jsou přímo využitelné v praxi. Tímto tedy začínáme publikovat v recenzované rubrice články typu metodiky, které procházejí posuzovacím procesem (v tomto případě v rámci MŽP, z něhož mají vyjít jako obecně doporučený materiál.V rubrice Informace máme ještě článek Petra Šauera (rekapituluje poslední inovace ve studiu environmentální ekonomie a Katky Jančaříkové (zpráva o konferenci Děti a zahrady, a několik aktuálních novinek, ale to je pro tentokrát vše, mimo jiné i proto, že naši recenzenti dělají svou práci pečlivě, jsou p