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Sample records for air primer pwr

  1. Detonation of hydrogen-air mixtures. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.H.S.; Knystautas, R.; Benedick, W.B.

    1983-01-01

    The detonation of a hydrogen-air cloud subsequent to an accidental release of hydrogen into ambient surroundings cannot be totally ruled out in view of the relative sensitivity of the hydrogen-air system. The present paper investigates the key parameters involved in hydrogen-air detonations and attempts to establish quantitative correlations between those that have important practical implications. Thus, for example, the characteristic length scale lambda describing the cellular structure of a detonation front is measured for a broad range of hydrogen-air mixtures and is quantitatively correlated with the key dynamic detonation properties such as detonability, transmission and initiation.

  2. Wind Energy and Air Emission Reduction Benefits: A Primer

    Energy Technology Data Exchange (ETDEWEB)

    Jacobson, D.; High, C.

    2008-02-01

    This document provides a summary of the impact of wind energy development on various air pollutants for a general audience. The core document addresses the key facts relating to the analysis of emission reductions from wind energy development. It is intended for use by a wide variety of parties with an interest in this issue, ranging from state environmental officials to renewable energy stakeholders. The appendices provide basic background information for the general reader, as well as detailed information for those seeking a more in-depth discussion of various topics.

  3. Oligo-cyclic damage and behaviour of a 304 L austenitic stainless steel according to environment (vacuum, air, PWR primary water) at 300 C

    International Nuclear Information System (INIS)

    Nowadays, for nuclear power plants licensing or operating life extensions, various safety authorities require the consideration of the primary water environment effect on the fatigue life of Pressurized Water Reactor (PWR) components. Thus, this work focused on the study of low cycle fatigue damage kinetics and mechanisms, of a type 304L austenitic stainless steel. Several parameters effects such as temperature, strain rate or strain amplitude were investigated in air as in PWR water. Thanks to targeted in-vacuum tests, the intrinsic influence of these parameters and environments on the fatigue behaviour of the material was studied. It appears that compared with vacuum, air is already an active environment which is responsible for a strong decrease in fatigue lifetime of this steel, especially at 300 C and low strain amplitude. The PWR water coolant environment is more active than air and leads to increased damage kinetics, without any modifications of the initiation sites or propagation modes. Moreover, the decreased fatigue life in PWR water is essentially attributed to an enhancement of both initiation and micropropagation of 'short cracks'. Finally, the deleterious influence of low strain rates on the 304L austenitic stainless steel fatigue lifetime was observed in PWR water environment, in air and also in vacuum without any environmental effects. This intrinsic strain rate effect is attributed to the occurrence of the Dynamic Strain Aging phenomenon which is responsible for a change in deformation modes and for an enhancement of cracks initiation. (author)

  4. Por mano propia. La justicia policial de la provincia de Buenos Aires en el primer peronismo

    Directory of Open Access Journals (Sweden)

    Osvaldo Barreneche

    2009-11-01

    Full Text Available La sanción del Código Penal Policial fue una pieza importante en el andamiaje jurídico por el cual el primer gobierno peronista procuró afianzar la profesionalización y disciplinamiento de las fuerzas de seguridad en las distintas jurisdicciones del país. Así, se organizó y se puso en marcha una nueva rama de la justicia bonaerense, paralela a la ordinaria, integrada completamente por policías retirados y en actividad, que se ocupó de los casos penales donde estuviesen implicados uno o más policías. Este trabajo analiza dicha experiencia, basándose en los casos sumariados y resueltos por la Justicia Policial mientras existió. En primer lugar se pone en contexto los fundamentos y propósitos que guiaron al peronismo para impulsar una justicia 'uniformada' que se ocupase de los excesos y falta a los deberes de los funcionarios policiales a nivel penal y administrativo. Luego se estudia su aplicación concreta en la provincia de Buenos Aires, indagando sobre los casos juzgados, su tratamiento, los actores, los fallos y los fines políticos perseguidos por esta justicia por y para policías

  5. Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-01

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

  6. Phonics Primer

    Science.gov (United States)

    Elam, Sandra

    2007-01-01

    This primer lists the 44 sounds in the English language and then gives steps for teaching those 44 sounds and their most common spelling patterns. In addition to learning sounds and spellings, each day the student must read lists of phonetically related words and spell these words from dictation. Phonics instruction must be reinforced by having…

  7. RATMAC PRIMER

    Energy Technology Data Exchange (ETDEWEB)

    Munn, R. J.; Stewart, J. M.; Norden, A. P.; Pagoaga, M. Katherine

    1980-10-01

    The language RATMAC is a direct descendant of one of the most successful structured FORTRAN languages, rational FORTRAN, RATFOR. RATMAC has all of the characteristics of RATFOR, but is augmented by a powerful recursive macro processor which is extremely useful in generating transportable FORTRAN programs. A macro is a collection of programming steps which are associated with a keyword. This keyword uniquely identifies the macro, and whenever it appears in a RATMAC program it is replaced by the collection of steps. This primer covers the language's control and decision structures, macros, file inclusion, symbolic constants, and error messages.

  8. Amorales, patoteros, chongos y pitucos. La homosexualidad masculina durante el primer peronismo [Buenos Aires, 1943-1955

    Directory of Open Access Journals (Sweden)

    Omar Acha

    2004-11-01

    Full Text Available Este artículo estudia la identificación de los homosexuales varones como un grupo singular durante la primera década peronista, bajo la figura de los "amorales". A diferencia de principios de siglo XX, donde estaban integrados en el caótico ambiente de la mala vida. Los varones de mediados de siglo que tenían relaciones sexuales con varones comenzaron a ser destacados como una desviación peligrosa del ideal familiarista y heterosexista. Este proceso se explica a través de la masculinización de la clase obrera operada por el peronismo y la aparición contemporánea de los jóvenes como un sector social problemático y peligroso. Se estudia este desarrollo de la intolerancia comenzando en los primeros años '40 hasta llegar a las razzias antihomosexuales de 1954-1955 en las calles, bares y plazas de Buenos Aires. A diferencia de otras interpretaciones que subrayan el carácter ancilar de las batidas policiales, sostenemos que fueron más que una excusa en el combate del peronismo contra el catolicismo, pues se inscribían en tensiones más profundas del orden social, sexual y simbólico de la Argentina de posguerra.

  9. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations

  10. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  11. PWR degraded core analysis

    International Nuclear Information System (INIS)

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  12. Application of the total etching technique or self-etching primers on primary teeth after air abrasion Aplicação da técnica de condicionamento total ou de "primers" autocondicionantes em dentes decíduos após abrasão a ar

    Directory of Open Access Journals (Sweden)

    Fábio Renato Manzolli Leite

    2005-09-01

    Full Text Available Since the use of air abrasion has grown in pediatric dentistry, the aim of this study was to evaluate, by means of shear bond strength testing, the need to use the total etching technique or self-etching primers on dentin of primary teeth after air abrasion. Twenty-five exfoliated primary molars had their occlusal dentin exposed by trimming and polishing. Specimens were treated by: Air abrasion + Scotchbond MultiPurpose adhesive (G1; 37% phosphoric acid + Scotchbond MP adhesive (G2; Clearfil SE (G3; Air abrasion + 37% phosphoric acid + Scotchbond MP adhesive (G4; Air abrasion + Clearfil SE (G5. On the treated surface, a cylinder of 2 mm by 6 mm was made using a composite resin (Z100. Duncan's test showed that: (G2 = G3 = G5 > (G1 = G4. The use of a self-etching primer on air abraded dentin is recommended to obtain higher bond strengths.Como o uso da abrasão a ar tem aumentado em odontopediatria, o objetivo deste trabalho foi avaliar a resistência ao cisalhamento da resina composta Z100 sobre a superfície dentinária de dentes decíduos após abrasionamento a ar associado ou não a técnica de condicionamento total (ácido fosfórico + Scotchbond Multi Purpose ou "primer" autocondicionante (Clearfil SE Bond. A superfície oclusal de 25 molares decíduos foi removida para exposição completa da superfície dentinária. Após polimento com lixas abrasivas, foram divididos em 5 grupos (5 espécimes em cada tratados por abrasão + adesivo Scotchbond (G1; condicionamento ácido + adesivo Scotchbond (G2; adesivo Clearfil (G3; abrasão a ar + condicionamento ácido + adesivo Scotchbond (G4; abrasão a ar + adesivo Clearfil (G5. Uma matriz bipartida foi adaptada à superfície dentinária e preenchida com resina composta. Após polimerização, corpos-de-prova de 2 mm x 6 mm foram obtidos. O teste de Duncan mostrou a desigualdade (G1 = G4 < (G2 = G3 = G5. Desse modo, conclui-se que, após o abrasionamento dentinário, deve ser utilizado o sistema de

  13. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  14. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  15. Physics of hydride fueled PWR

    Science.gov (United States)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  16. Estrategia de marketing para la comercialización de servicios de telefonía IP para llamadas internacionales en la población extranjera que vive en el primer cordón del Gran Buenos Aires

    OpenAIRE

    Joya Castellanos, Carlos Alberto

    2014-01-01

    Al igual que la telefonía rural en los países en vía de desarrollo, donde las compañías operadoras de telefonía no invierten porque la demanda de abonados no es suficiente para cubrir los estándares de equipamiento y operación y el costo de mantenimiento es muy alto por el desplazamiento del capital de trabajo a zonas alejadas. En el primer cordón del Gran Buenos Aires, existe una situación similar, en la que las principales compañías operadoras de telefonía de la Argentina, no hacen inversió...

  17. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    The authors believe the next generation nuclear power plant should be characterized by: (1) simplicity of design; (2) ease of operation and maintenance; (3) economic conformance with safety requirements; and (4) technologies easy to understand by the public. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, the Japan Atomic Power Company (JAPC) supported by the other Japanese PWR Utilities, Electricite de France (EdF), Westinghouse (WH) and Mitsubishi Heavy Industry (MHI) have studied application of passive technologies at a power rating of about 1,000 MWe. The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Using the AP-600 reference design as a basis, the authors enlarged the plant size to 3 -loops and added engineering features to conform with Japanese practice and Utilities' preference. The Simplified PWR (SPWR) program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1,000 MWe and 3 loops

  18. Uruguayos en Buenos Aires: Procesos Sociales de Marcación, Trabajos de Legitimación y Desigualdad entre el Primer Peronismo y las Papeleras

    Directory of Open Access Journals (Sweden)

    Merenson Silvina

    2014-12-01

    Full Text Available Si bien no puede afirmarse que los uruguayos en BuenosAires forman parte de los “flujos migratorios deseables”, puede advertirse los esfuerzos de este colectivo por no quedar asimilados al “talón de Aquiles del crisol de razas”. A partir de un trabajo etnográfico multi-situado, este artículo aborda los procesos sociales demarcación y los trabajos de justificación emprendidos por distintos uruguayos residentes en Buenos Aires para distinguir sus trayectorias migratorias de otras posibles. Dichos procesos y trabajos son analizados en relación con los sentidos y prácticas asociadas a lo que este colectivo denomina “hermandad rioplatense” – así como con sus límites – entremediados de la década de 1940 y la actualidad. En definitiva, este artículo propone analizar estas operaciones a partir de las interdependencias entre las activaciones de determinadas narrativas nacionales, las relaciones bilaterales entre Argentina y Uruguay y las desigualdades estructurales en la región.

  19. Educación y obra pública durante el primer peronismo : La construcción de escuelas en la provincia de Buenos Aires

    OpenAIRE

    Petitti, Eva Mara

    2016-01-01

    Este artículo se propone analizar el Plan Integral de Edificación Escolar (PIEE) que se aprobó en la provincia de Buenos Aires en 1948. El estudio de los cambios que implicó en relación a los proyectos anteriores, y de los alcances y límites de su aplicación durante las gestiones de Domingo A. Mercante (1946-1952) y Carlos Aloé (1952-1955), permite señalar que la ejecución del plan de edificación que impactó de manera significativa en el número de establecimientos educativos fiscales, fue pos...

  20. Educación y obra pública durante el primer peronismo. La construcción de escuelas en la provincia de Buenos Aires

    OpenAIRE

    Eva Mara Petitti

    2016-01-01

    Este artículo se propone analizar el Plan Integral de Edificación Escolar (PIEE) que se aprobó en la provincia de Buenos Aires en 1948. El estudio de los cambios que implicó en relación a los proyectos anteriores, y de los alcances y límites de su aplicación durante las gestiones de Domingo A. Mercante (1946-1952) y Carlos Aloé (1952-1955), permite señalar que la ejecución del plan de edificación que impactó de manera significativa en el número de establecimientos educativos fiscales, fue pos...

  1. China Energy Primer

    Energy Technology Data Exchange (ETDEWEB)

    Ni, Chun Chun

    2009-11-16

    Based on extensive analysis of the 'China Energy Databook Version 7' (October 2008) this Primer for China's Energy Industry draws a broad picture of China's energy industry with the two goals of helping users read and interpret the data presented in the 'China Energy Databook' and understand the historical evolution of China's energy inustry. Primer provides comprehensive historical reviews of China's energy industry including its supply and demand, exports and imports, investments, environment, and most importantly, its complicated pricing system, a key element in the analysis of China's energy sector.

  2. Carta dinámica del medio ambiente del Partido de Luján, provincia de Buenos Aires-República Argentina. Primer etapa

    Directory of Open Access Journals (Sweden)

    Nelly M. Guadalupe Leclerc

    1996-01-01

    Full Text Available La Carta Dinámica del Medio Ambiente es un instrumento que p e d , a través de la evaluación de diversos parámetros, Uevar a cabo diagnóstiws ambientales En la República Argatina son pocas las experiencias llevadas a cabo al respecto, en este caso particular resulta de sumo interés desanollar la Carta D i c a &&da a un Partido del área de inüuencia de la Universidad Nacional de Luján, smiado al oeste de la ciudad de Buenos Aires (aproximadamente 70 km, en la República -tina. Para este estudio se makwá un detallado análisis de diversos componentes del medio: topografía, &m, hidrografia, suelos, usos de la tierra, etc., y se tiene en cuenta asimismo la resultante de sus wmbiiiones: degradacih, erosión, contamhación del suelo, aire y agua, entre otros. Para representar toda la información obtenida se elaborarán dos cartas a escala 150 000; en la Carta 1 se voicarán los dafos wrrespdientes a los elementos del medio y en la Carta 2 el Estado de los mismos. Se empleará la súnbología eskwdarkda por la Unión M c a Internacional (UGI para cartas dinámicas del medio ambiente. Se aplicará procesamiento &&al, nivelación de histogramas y realce de bordes sobre la imagen LANDSAT TM 225-084 de 1992, para logm un producto especial que pemUtirá realizar interpretación visual y con ello la posterior obtención de distintos mapas temáticos. También se bizo uso de cartas topogmfícas, fotcgrañas aéreas, datos del Servicio Meteorológico Nacional y otros. Toda la información se i n t m en una base de datos mediante la uiüización del sistema de infonnaciá geográfica (SIG ARC-WFO. Aquí se presenta la primera etapa de este trabajo en donde se pmcedió al análisis de las variables relativas a la topograña, red hidrográiica, vias de wmunicacion, industrias y centros urbanos, obteniéndose wmo resultado mapas p r e h h m para la elaboración de la Carta 1.

  3. An SAT® Validity Primer

    Science.gov (United States)

    Shaw, Emily J.

    2015-01-01

    This primer should provide the reader with a deeper understanding of the concept of test validity and will present the recent available validity evidence on the relationship between SAT® scores and important college outcomes. In addition, the content examined on the SAT will be discussed as well as the fundamental attention paid to the fairness of…

  4. Coal Bed Methane Primer

    Energy Technology Data Exchange (ETDEWEB)

    Dan Arthur; Bruce Langhus; Jon Seekins

    2005-05-25

    During the second half of the 1990's Coal Bed Methane (CBM) production increased dramatically nationwide to represent a significant new source of income and natural gas for many independent and established producers. Matching these soaring production rates during this period was a heightened public awareness of environmental concerns. These concerns left unexplained and under-addressed have created a significant growth in public involvement generating literally thousands of unfocused project comments for various regional NEPA efforts resulting in the delayed development of public and fee lands. The accelerating interest in CBM development coupled to the growth in public involvement has prompted the conceptualization of this project for the development of a CBM Primer. The Primer is designed to serve as a summary document, which introduces and encapsulates information pertinent to the development of Coal Bed Methane (CBM), including focused discussions of coal deposits, methane as a natural formed gas, split mineral estates, development techniques, operational issues, producing methods, applicable regulatory frameworks, land and resource management, mitigation measures, preparation of project plans, data availability, Indian Trust issues and relevant environmental technologies. An important aspect of gaining access to federal, state, tribal, or fee lands involves education of a broad array of stakeholders, including land and mineral owners, regulators, conservationists, tribal governments, special interest groups, and numerous others that could be impacted by the development of coal bed methane. Perhaps the most crucial aspect of successfully developing CBM resources is stakeholder education. Currently, an inconsistent picture of CBM exists. There is a significant lack of understanding on the parts of nearly all stakeholders, including industry, government, special interest groups, and land owners. It is envisioned the Primer would being used by a variety of

  5. Primer on molecular genetics

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This report is taken from the April 1992 draft of the DOE Human Genome 1991--1992 Program Report, which is expected to be published in May 1992. The primer is intended to be an introduction to basic principles of molecular genetics pertaining to the genome project. The material contained herein is not final and may be incomplete. Techniques of genetic mapping and DNA sequencing are described.

  6. Activity transport models for PWR primary circuits

    International Nuclear Information System (INIS)

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR's. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.)

  7. Program of monitoring PWR fuel in Spain

    International Nuclear Information System (INIS)

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  8. Math primer for engineers

    CERN Document Server

    Cryer, CW

    2014-01-01

    Mathematics and engineering are inevitably interrelated, and this interaction will steadily increase as the use of mathematical modelling grows. Although mathematicians and engineers often misunderstand one another, their basic approach is quite similar, as is the historical development of their respective disciplines. The purpose of this Math Primer is to provide a brief introduction to those parts of mathematics which are, or could be, useful in engineering, especially bioengineering. The aim is to summarize the ideas covered in each subject area without going into exhaustive detail. Formula

  9. The R primer

    CERN Document Server

    Ekstrom, Claus Thorn

    2011-01-01

    Newcomers to R are often intimidated by the command-line interface, the vast number of functions and packages, or the processes of importing data and performing a simple statistical analysis. The R Primer provides a collection of concise examples and solutions to R problems frequently encountered by new users of this statistical software.Rather than explore the many options available for every command as well as the ever-increasing number of packages, the book focuses on the basics of data preparation and analysis and gives examples that can be used as a starting point. The numerous examples i

  10. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  11. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.)

  12. Numerical simulation using CFD software of countercurrent gas-liquid flow in a PWR hot leg under reflux condensation

    Energy Technology Data Exchange (ETDEWEB)

    Utanohara, Yoichi, E-mail: utanohara@inss.co.j [Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 919-1205 (Japan); Kinoshita, Ikuo, E-mail: kinoshita@inss.co.j [Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 919-1205 (Japan); Murase, Michio, E-mail: murase@inss.co.j [Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 919-1205 (Japan); Minami, Noritoshi, E-mail: minami.noritoshi@c4.kepco.co.j [Kansai Electric Power Company, Inc., Mihama-cho, Mikata-gun, Fukui 919-1141 (Japan); Nariai, Toshifumi [Graduate School of Engineering, Kobe University, Nada, Kobe 657-8501 (Japan); Tomiyama, Akio, E-mail: tomiyama@people.kobe-u.ac.j [Graduate School of Engineering, Kobe University, Nada, Kobe 657-8501 (Japan)

    2011-05-15

    Research highlights: The wave height measurements indicated the interfacial drag force mainly consisted of form drag. The resolution of the computational cell affected the CCFL characteristics. The interfacial drag correlations employed in this study can be applied to PWR scale simulations. - Abstract: In order to improve the countercurrent flow model of a transient analysis code, countercurrent air-water tests were previously conducted using a 1/15 scale model of the PWR hot leg and numerical simulations of the tests were carried out using the two-fluid model implemented in the CFD software FLUENT 6.3.26. The predicted flow patterns and CCFL characteristics agreed well with the experimental data. However, the validation of the interfacial drag correlation used in the two-fluid model was still insufficient, especially regarding the applicability to actual PWR conditions. In this study, we measured water levels and wave heights in the 1/15 scale setup to understand the characteristics of the interfacial drag, and we considered a relationship between the wave height and the interfacial drag coefficient. Numerical simulations to examine the effects of cell size and interfacial drag correlations on numerical predictions were conducted under PWR plant conditions. Wave heights strongly related with the water level and interfacial drag coefficient, which indicates that the interfacial drag force mainly consists of form drag. The cell size affected the gas velocity at the onset of flooding in the process of increasing gas flow rate. The gas volumetric fluxes at CCFL predicted using fine cells were higher than those using normal cells. On the other hand, the cell size did not have a significant influence on the process of decreasing gas flow rate. The predictions for the PWR condition using a reference set of interfacial drag correlations agreed well with the Upper Plenum Test Facility data of the PWR scale experiment in the region of medium gas volumetric fluxes. The

  13. PWR-blowdown heat transfer separate effects program

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described.

  14. Frictional Behavior of Fe-based Cladding Candidates for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Hyung-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byun, Thak Sang [Oak Ridge National Lab., Oak Ridge (United States)

    2014-10-15

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  15. Crack growth rates of nickel alloy welds in a PWR environment.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.; Energy Technology

    2006-05-31

    In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320 C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of {approx}5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking.

  16. Raanan Rein, Carolina Barry, Nicolás Quiroga y Omar Acha, Los estudios sobre el primer peronismo: aproximaciones desde el siglo XXI, La Plata, Instituto Cultural de la Provincia de Buenos Aires, 2009, 118 p.

    OpenAIRE

    Reclusa, Alejo

    2010-01-01

    Esta compilación surge como resumen de lo discutido en el “Primer Congreso de Estudios sobre el Peronismo: La Primera Década”, realizado en la Universidad Nacional de Mar del Plata los días 6 y 7 de noviembre de 2008. El libro consta de tres artículos: el primero a cargo del historiador israelí Raanan Rein, que intenta un relevamiento historiográfico sobre el estudio del primer peronismo; el segundo artículo a cargo de Carolina Barry, centrado en los aportes de las nuevas perspectivas y la co...

  17. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  18. Thorium fuel cycle study for PWR applications

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae Yong; Kim, Myung Hyun [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-12-31

    A nuclear design feasibility of thorium fueled high converting PWR was investigated. Two kinds of fuel design option were tested for the comparison with conventional UO{sub 2} fuel. The first one was an application of MHTGR pyro-carbon coated particle fuels. The other design was an application of MOX fuels as a ThO{sub 2}-PuO{sub 2} ceramic pellet. In the case of carbon-coated particle fuels, there was no benefit in nuclear design aspect because enrichment of U-235 was required over 5 w/o in order to match with the K-infinite of Ulchin-3/4 fuels. However, the use of thorium based plutonium fuels in PWR gave favorable aspects in nuclear design such as flatter K-infinite curve, lower M. T. C. and lower F. T. C. than that of UO{sub 2} fuel. (author). 6 refs., 3 tabs., 6 figs.

  19. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  20. The integrated PWR; Les REP integres

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2002-07-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  1. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  2. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  3. 百万级压水堆核电站空冷汽轮机选型研究%Calculation and Selecting of Air-cooled Turbine in 1 000 MW PWR Nuclear Power Station

    Institute of Scientific and Technical Information of China (English)

    杨建军; 孙传军; 赵迪

    2015-01-01

    针对国内某核电厂址的初步气象条件,进行了百万等级核电机组空冷汽轮机选型计算分析。分析结果表明,该厂址参数下若选用半转速核电汽轮机,目前已有的末级叶片不能满足要求,需要开发全新的空冷末级叶片;而利用大型火电空冷汽轮机已有运行业绩的末级叶片,开发全转速百万级核电空冷汽轮机,将花费较少的时间和经费,是比较适宜的方案。某厂址气象条件在我国北方具有代表性,上述工作为我国北方缺水地区核电站汽轮机选型提供了参考。建议国内汽轮机厂利用已有核电、火电汽轮机技术积累,尽快启动全转速百万级核电空冷汽轮机研发工作,形成具有完全自主知识产权的百万级核电空冷汽轮机技术,抢占技术制高点。%By analyzing the weather characters of one potential 1000MW nuclear power station site ,this paper calculated the air-cooled steam turbine exhausting area .It is concluded that the present last long blade used in half-speed turbine is not suitable,new last long blade shall be developed if we want to use half-speed turbine in 1 000MW air-cooled nuclear power station.However,the full-speed turbines and their last long blades used in coal burning power station are appropriate for this site,excepting plus one low pressure turbine and some measures to remove moisture steam in the last stage .The latter is better,for it will cost less money and time .The site mentioned in this paper is representive in China north area ,and this paper is useful for turbine selecting of nuclear power station in this area .Domestic turbine manufacturers have accumulated many experiences and technologies by manufacturing and developing half -speed nuclear turbine and air-cooled full-speed turbines in coal burning power station .It is suggested the domestic turbine manufacturers should setup the research work as soon as possible .

  4. Effect of surface finish and loading conditions on the LCF behavior of austenitic stainless steel in PWR environment

    International Nuclear Information System (INIS)

    ANL has issued in 2006 a NUREG/CR-6909 report that is now applicable in the US for evaluations of PWR environmental effects in the fatigue analysis of new reactor components. In order to assess the conservativeness of the application of this NUREG report, low cycle fatigue (LCF) tests were performed by AREVA NP on austenitic stainless steel specimens in a PWR environment. The selected material exhibits in an air environment a fatigue behavior consistent with the ANL reference 'air' mean curve. Tests were performed in PWR environment for two various loading conditions: for fully reverse triangular signal (for comparison purpose with tests performed by other laboratories with same loading conditions) and complex signal, simulating strain variation for actual typical PWR thermal transients. Two surface finish conditions were tested: polished and ground. This paper presents the comparison of environmental penalty factors (Fen) as observed experimentally with the ANL formulation (considering the strain integral method for complex loading), and the actual fatigue life of the specimen with the fatigue life predicted through the NUREG/CR-6909 application

  5. Air

    Science.gov (United States)

    ... house) Industrial emissions (like smoke and chemicals from factories) Household cleaners (spray cleaners, air fresheners) Car emissions (like carbon monoxide) *All of these things make up “particle pollution.” They mostly come from cars, trucks, buses, and ...

  6. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  7. Air

    International Nuclear Information System (INIS)

    In recent years several regulations and standards for air quality and limits for air pollution were issued or are in preparation by the European Union, which have severe influence on the environmental monitoring and legislation in Austria. This chapter of the environmental control report of Austria gives an overview about the legal situation of air pollution control in the European Union and in specific the legal situation in Austria. It gives a comprehensive inventory of air pollution measurements for the whole area of Austria of total suspended particulates, ozone, volatile organic compounds, nitrogen oxides, sulfur dioxide, carbon monoxide, heavy metals, benzene, dioxin, polycyclic aromatic hydrocarbons and eutrophication. For each of these pollutants the measured emission values throughout Austria are given in tables and geographical charts, the environmental impact is discussed, statistical data and time series of the emission sources are given and legal regulations and measures for an effective environmental pollution control are discussed. In particular the impact of fossil-fuel power plants on the air pollution is analyzed. (a.n.)

  8. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  9. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  10. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  11. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing keff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR keff markedly. The PWR keff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  12. Vygotsky on Education Primer. Peter Lang Primer. Volume 30

    Science.gov (United States)

    Lake, Robert

    2012-01-01

    The "Vygotsky on Education Primer" serves as an introduction to the life and work of the Russian psychologist Lev Vygotsky. Even though he died almost eighty years ago, his life's work remains both relevant and significant to the field of education today. This book examines Vygotsky's emphasis on the role of cultural and historical context in…

  13. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.)

  14. Study of safety relief valve operation under ATWS conditions. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  15. Development of a dry storage cask for PWR spent fuel

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co., Ltd.(KHNP), which operates all the nuclear power plants in Korea, is developing a new dry storage cask to store twenty four spent fuel assemblies generated from pressurized water reactors for at-reactor or away-from-reactor interim storage facility in Korea. The dry storage cask is designed and evaluated according to the requirements of the IAEA, the US NRC and the Korean regulations for the dry spent fuel storage system. It provides confinement, radiation shielding, structural integrity, subcritical control and passive heat removal for normal and accident conditions. The dry storage cask consists of a dual purpose canister providing a confinement boundary for the PWR spent fuel, and a storage overpack providing a structural and radiological boundary for long-term storage of the canister placed inside it. The overpack is constructed by a combination of steel and concrete, and is equipped with penetrating ducts near its lower and upper extremities to permit natural circulation of air to provide for the passive cooling of the canister and the contained spent fuel assemblies. This paper describes development status, description, design criteria, evaluation and demonstration tests of the dry storage cask. (authors)

  16. A Hearing Aid Primer 1

    Science.gov (United States)

    Yetter, Carol J.

    2009-01-01

    This hearing aid primer is designed to define the differences among the three levels of hearing instrument technology: conventional analog circuit technology (most basic), digitally programmable/analog circuit technology (moderately advanced), and fully digital technology (most advanced). Both moderate and advanced technologies mean that hearing…

  17. DNA Extraction and Primer Selection

    DEFF Research Database (Denmark)

    Karst, Søren Michael; Nielsen, Per Halkjær; Albertsen, Mads;

    Talk regarding pitfalls in DNA extraction and 16S amplicon primer choice when performing community analysis of complex microbial communities. The talk was a part of Workshop 2 "Principles, Potential, and Limitations of Novel Molecular Methods in Water Engineering; from Amplicon Sequencing to -omics...

  18. Freshwater Wetlands: A Citizen's Primer.

    Science.gov (United States)

    Catskill Center for Conservation and Development, Inc., Hobart, NY.

    The purpose of this "primer" for the general public is to describe the general characteristics of wetlands and how wetland alteration adversely affects the well-being of humans. Particular emphasis is placed on wetlands in New York State and the northeast. Topics discussed include wetland values, destruction of wetlands, the costs of wetland…

  19. A classical primer for QCD

    International Nuclear Information System (INIS)

    A basic primer for QCD is presented using a semiclassical approach to the colour Maxwell equations. The non-Abelian nature of colour symmetry and the violation of superposition by colour fields is compared with QED. A simple discussion of asymptotic freedom is also presented. (author)

  20. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  1. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  2. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  3. Crevice chemistry control in PWR steam generators

    International Nuclear Information System (INIS)

    To establish a basis for predicting and eventually controlling crevice solution chemistry in PWR steam generators, hideout tests were performed at several units. Results indicated that impurity hideout rates varied with the species and with bulk water concentration. Field evaluations of crevice impurity inventory models based on the hideout rate data indicated that further model refinements were necessary, e.g., more frequent quantification of the relation of hideout rates and bulk water concentration. An alternate crevice inventory model based on a real-time mass balance approach also began to be pursued. Modeling results currently are being used at several PWRs to establish a chloride injection rate consistent with development of a near neutral crevice solution to minimize IGA/SCC. Hideout return data are being used to independently establish predictions of crevice chemistry and to substantiate the hideout rate and mass balance model predictions

  4. The underclad cracking in PWR reactor vessels

    International Nuclear Information System (INIS)

    The article describes the kind of cracking which can occur under the stainless steel cladding during the manufacturing process of PWR vessels: - cold cracking recently found in France on vessel nozzles-reheat cracking discovered some ten years ago in particular in Germany and in USA. Methods of examination for underclad cracking are put forward, together with results obtained on vessel nozzles of units currently being built in Belgium. Some nozzles are affected by the phenomenon of reheat cracking, whilst the hypothesis of cold cracking, which had been proposed because of the similar situation found in France should probably be abandoned. On the basis of the investigations and studies made, it is established that the cracking involved does not jeopardize the integrity of the vessels during their life time. (author)

  5. The material analysis for PWR primary equipment

    International Nuclear Information System (INIS)

    The primary equipment in pressurized water reactor includes reactor pressure vessel, reactor coolant piping, steam generator, pressurizer, and reactor coolant pump casing, etc., which form the pressure boundary of the primary loop. These primary equipment are all pressure vessels of QA Class 1, Safety-related Class 1, and Aseismatic Category 1. Under high temperature, high pressure and neutron irradiation, the requirements for the base material and welding properties of these pressure vessels are very high, so as to ensure the long-term stable operation of nuclear power plant. The base material and welding properties of these pressure vessels are analyzed and discussed according to ASME B and P Code, which can be as a reference for base material selection of PWR pressure vessels. (authors)

  6. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  7. The PWR cores management; La gestion des coeurs REP

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Rippert, D. [CEA Cadarache, Departement d' Etudes des Reacteurs, DER, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    2000-01-25

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  8. Hydraulic benchmark data for PWR mixing vane grid

    International Nuclear Information System (INIS)

    The purpose of the present study is to present new hydraulic benchmark data obtained for PWR rod bundles for the purpose of benchmarking Computational Fluid Dynamics (CFD) models of the rod bundle. The flow field in a PWR fuel assembly downstream of structural grids which have mixing vane grids attached is very complex due to the geometry of the subchannel and the high axial component of the velocity field relative to the secondary flows which are used to enhance the heat transfer performance of the rod bundle. Westinghouse has a CFD methodology to model PWR rod bundles that was developed with prior benchmark test data. As improvements in testing techniques have become available, further PWR rod bundle testing is being performed to obtain advanced data which has high spatial and temporal resolution. This paper presents the advanced testing and benchmark data that has been obtained by Westinghouse through collaboration with Texas A&M University. (author)

  9. Hot Operation of FTL for PWR Fuels Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sung Ho; Joung, Chang Yong; Lee, Jong Min; Park, Su Ki; Sim, Bong Sik; Ahn, Guk Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Fuel Test Loop (FTL) in HANARO is the test facility which can conduct a fuel irradiation test with commercial NPPs' operating conditions such as their pressure, temperature, flow and water chemistry. The FTL is used for the irradiation test of PWR type or CNNDU type fuels. In this paper, the hot operation of FTL for irradiation test of PWR fuels is introduced. The experimental results show the excellence of operation performance

  10. 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)

  11. Seawater desalination using reusable type small PWR

    Energy Technology Data Exchange (ETDEWEB)

    Uchiyama, Y. [Institute of Engineering Mechanics and Systems, University of Tsukuba, Tsukuba, Ibaraki (Japan); Minato, A. [Planning Division, Central Research Institute of the Electric Power Industry, Komae-shi, Tokyo (Japan); Shimamura, K. [Nuclear Systems Engineering Department, Nuclear Energy Systems Engineering Center, Mitsubishi Heavy Industries, Ltd., Kanagawa (Japan)]. E-mail: shimamura@atom.hq.mhi.co.jp

    2003-07-01

    Demand for seawater desalination is increasing, especially in regions such as the Middle East and North Africa, where populations are growing at a high annual rate. If such demand is met by fossil fuel energy, the influence on the environment, such as global warming, cannot be disregarded. Since these regions are behind in their preparedness of social capital infrastructure, such as power transfer grids, small reactors are considered to be more suitable for introduction than the large reactors found commonly in developed countries. Therefore, a small reusable PWR with mid-range pressure and temperature services, which does not require on-site refuelling, was devised for seawater desalination. In a small reusable PWR, spent fuel is taken out together with the reactor vessel and refuelled on the exterior fuel exchange base prepared independently. Thus, the safeguards against nuclear proliferation increase at a plant site because the lid of the reactor vessel is never opened at the site, in principle. The reactor vessel will be transported from the plant site to a fuel exchange base under stipulated conditions within a transportation cask after a long (about six years) operation. Since fuel handling facilities at the site become unnecessary through centralisation at a fuel exchange base, initial plant construction costs are reduced. In addition, the reactor vessel is reused until its service life has expired. This examination was based on the marine reactor of the experimental nuclear ship, Mutsu, after it had been applied for land use: at a lowered, midrange pressure and temperature service, in theory. It is possible to produce fresh water through reverse osmosis (RO) membrane pressure-rising seawater by a steam turbine driven pump. Using the method of driving a desalination unit high-pressure pump directly by low-pressure steam generated from the heating reactor, fresh water can be produced efficiently. Furthermore, operating at reduced pressure makes it possible

  12. The Study of Nuclear Fuel Cycle Options Based On PWR and CANDU Reactors

    International Nuclear Information System (INIS)

    The study of nuclear fuel cycle options based on PWR and CANDU type reactors have been carried out. There are 5 cycle options based on PWR and CANDU reactors, i.e.: PWR-OT, PWR-OT, PWR-MOX, CANDU-OT, DUPIC, and PWR-CANDU-OT options. While parameters which assessed in this study are fuel requirement, generating waste and plutonium from each cycle options. From the study found that the amount of fuel in the DUPIC option needs relatively small compared the other options. From the view of total radioactive waste generated from the cycles, PWR-MOX generate the smallest amount of waste, but produce twice of high level waste than DUPIC option. For total plutonium generated from the cycle, PWR-MOX option generates smallest quantity, but for fissile plutonium, DUPIC options produce the smallest one. It means that the DUPIC option has some benefits in plutonium consumption aspects. (author)

  13. A primer of Lebesgue integration

    CERN Document Server

    Bear, H S

    2001-01-01

    The Lebesgue integral is now standard for both applications and advanced mathematics. This books starts with a review of the familiar calculus integral and then constructs the Lebesgue integral from the ground up using the same ideas. A Primer of Lebesgue Integration has been used successfully both in the classroom and for individual study.Bear presents a clear and simple introduction for those intent on further study in higher mathematics. Additionally, this book serves as a refresher providing new insight for those in the field. The author writes with an engaging, commonsense style that appeals to readers at all levels.

  14. Loop Quantum Geometry: A primer

    CERN Document Server

    Corichi, A

    2005-01-01

    This is the written version of a lecture given at the ``VI Mexican School of Gravitation and Mathematical Physics" (Nov 21-27, 2004, Playa del Carmen, Mexico), introducing the basics of Loop Quantum Geometry. The purpose of the written contribution is to provide a Primer version, that is, a first entry into Loop Quantum Gravity and to present at the same time a friendly guide to the existing pedagogical literature on the subject. This account is geared towards graduate students and non-experts interested in learning the basics of the subject.

  15. A primer of multivariate statistics

    CERN Document Server

    Harris, Richard J

    2014-01-01

    Drawing upon more than 30 years of experience in working with statistics, Dr. Richard J. Harris has updated A Primer of Multivariate Statistics to provide a model of balance between how-to and why. This classic text covers multivariate techniques with a taste of latent variable approaches. Throughout the book there is a focus on the importance of describing and testing one's interpretations of the emergent variables that are produced by multivariate analysis. This edition retains its conversational writing style while focusing on classical techniques. The book gives the reader a feel for why

  16. A primer of special relativity

    CERN Document Server

    Sardesai, PL

    2004-01-01

    A Primer of Special Relativity1 is an unusually lucid introduction to the subject specifically written for Indian students. It is intended to give the beginner a firm grounding for a more advanced course in relativity. An entire chapter is devoted to applications of the theory to elucidate a large number of topics the students (B.Sc. Physics) come across in Modern Physics. Detailed and well-selected examples are used to illuminate aspects of the theory as well as to show techniques of application. A large number of Illustrative Examples enables the students to gain confidence to solve any problem in relativity normally expected of B.Sc. students.

  17. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  18. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    The recent technologies of test, relaxation of deterioration, repairing and change of materials are explained for safe and stable operation of pressurized water reactor (PWR). Stress corrosion cracking (SCC) is originated by three factors such as materials, stress and environment. The eddy current test (ECT) method for the stream generator pipe and the ultrasonic test method for welding part of pipe were developed as the test technologies. Primary water stress corrosion cracking (PWSCC) of Inconel 600 in the welding part is explained. The shot peening of instrument in the gas, the water jet peening of it in water, and laser irradiation on the surface are illustrated as some examples of improvement technology of stress. The cladding of Inconel 690 on Inconel 600 is carried out under the condition of environmental cut. Total or some parts of the upper part of reactor, stream generator and structure in the reactor are changed by the improvement technologies. Changing Inconel 600 joint in the exit pipe of reactor with Inconel 690 is illustrated. (S.Y.)

  19. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

  20. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  1. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  2. UniPrimer: A Web-Based Primer Design Tool for Comparative Analyses of Primate Genomes

    Directory of Open Access Journals (Sweden)

    Nomin Batnyam

    2012-01-01

    Full Text Available Whole genome sequences of various primates have been released due to advanced DNA-sequencing technology. A combination of computational data mining and the polymerase chain reaction (PCR assay to validate the data is an excellent method for conducting comparative genomics. Thus, designing primers for PCR is an essential procedure for a comparative analysis of primate genomes. Here, we developed and introduced UniPrimer for use in those studies. UniPrimer is a web-based tool that designs PCR- and DNA-sequencing primers. It compares the sequences from six different primates (human, chimpanzee, gorilla, orangutan, gibbon, and rhesus macaque and designs primers on the conserved region across species. UniPrimer is linked to RepeatMasker, Primer3Plus, and OligoCalc softwares to produce primers with high accuracy and UCSC In-Silico PCR to confirm whether the designed primers work. To test the performance of UniPrimer, we designed primers on sample sequences using UniPrimer and manually designed primers for the same sequences. The comparison of the two processes showed that UniPrimer was more effective than manual work in terms of saving time and reducing errors.

  3. UniPrimer: A Web-Based Primer Design Tool for Comparative Analyses of Primate Genomes.

    Science.gov (United States)

    Batnyam, Nomin; Lee, Jimin; Lee, Jungnam; Hong, Seung Bok; Oh, Sejong; Han, Kyudong

    2012-01-01

    Whole genome sequences of various primates have been released due to advanced DNA-sequencing technology. A combination of computational data mining and the polymerase chain reaction (PCR) assay to validate the data is an excellent method for conducting comparative genomics. Thus, designing primers for PCR is an essential procedure for a comparative analysis of primate genomes. Here, we developed and introduced UniPrimer for use in those studies. UniPrimer is a web-based tool that designs PCR- and DNA-sequencing primers. It compares the sequences from six different primates (human, chimpanzee, gorilla, orangutan, gibbon, and rhesus macaque) and designs primers on the conserved region across species. UniPrimer is linked to RepeatMasker, Primer3Plus, and OligoCalc softwares to produce primers with high accuracy and UCSC In-Silico PCR to confirm whether the designed primers work. To test the performance of UniPrimer, we designed primers on sample sequences using UniPrimer and manually designed primers for the same sequences. The comparison of the two processes showed that UniPrimer was more effective than manual work in terms of saving time and reducing errors. PMID:22693428

  4. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10−6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  5. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14

  6. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  7. Single-primer fluorescent sequencing

    Energy Technology Data Exchange (ETDEWEB)

    Ruth, J.L.; Morgan, C.A.; Middendorf, L.R.; Grone, D.L.; Brumbaugh, J.A.

    1987-05-01

    Modified linker arm oligonucleotides complementary to standard M13 priming sites were synthesized, labelled with either one, two, or three fluoresceins, and purified by reverse-phase HPLC. When used as primers in standard dideoxy M13 sequencing with /sup 32/P-dNTPs, normal autoradiographic patterns were obtained. To eliminate the radioactivity, direct on-line fluorescence detection was achieved by the use of a scanning 10 mW Argon laser emitting 488 nm light. Fluorescent bands were detected directly in standard 0.2 or 0.35 mm thick polyacrylamide gels at a distance of 24 cm from the loading wells by a photomultiplier tube filtered at 520 nm. Horizontal and temporal location of each band was displayed by computer as a band in real time, providing visual appearance similar to normal 4-lane autoradiograms. Using a single primer labelled with two fluoresceins, sequences of between 500 and 600 bases have been read in a single loading with better than 98% accuracy; up to 400 bases can be read reproducibly with no errors. More than 50 sequences have been determined by this method. This approach requires only 1-2 ug of cloned template, and produces continuous sequence data at about one band per minute.

  8. Environmental effect on cracking of an 304L austenitic stainless steels in PWR primary environment under cyclic loading

    International Nuclear Information System (INIS)

    The present study was undertaken in order to get further insights on cracking mechanisms in a 304L stainless steel. More precisely, a first objective of this study was to evaluate the effect of various cold working conditions on the cyclic stress-strain behavior and the fatigue life in air and in PWR primary environment. In air a prior hardening was found to reduce the fatigue life in the LCF regime but not in primary environment. In both environments, the fatigue limit of the hardened materials was increased after cold working.The second objective addresses the effect of the air and the PWR primary environments on the cracking mechanisms (initiation and propagation) in the annealed material in the LCF regime. More precisely, the kinetics of crack initiation and micro crack propagation were evaluated with a multi scale microscopic approach in air and in primary environment. In PWR primary environment, during the first cycles, preferential oxidation occurs along emerging dissociated dislocation and each cycle generates a new C-rich/Fe-rich oxide layer. Then, during cycling, the microstructure evolves from stacking fault into micro twinning and preferential oxidation occurs by continuous shearing and dissolution of the passive film. Beyond a certain crack depth (≤3 μm), the crack starts to propagate with a direction close to a 90 degrees angle from the surface. The crack continues its propagation by successive generation of shear bands and fatigue striations at each cycle up to failure. The role of corrosion hydrogen on these processes is finally discussed. (author)

  9. Experimental investigation of reflux condensation heat transfer in PWR steam generator tubes in the presence of noncondensible gases

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen; Wu, Tiejun [Purdue Univ., West Lafayette (United States); Nagae, Takashi [Institute of Nuclear Safety System, Tokyo (Japan)

    2003-07-01

    Under certain circumstances in a Pressurized Water Reactor (PWR), the coolant system may be in a partially drained state and reflux condensation in the steam generator U-tubes can be the major heat removal mechanism. Noncondensable gases may be present and would degrade the heat transfer rate. If heat removal rates are insufficient, this situation could lead to core boil-off, fuel rod heatup, and eventually core damage. The Institute of Nuclear Safety System, Inc. (INSS) and the Nuclear Heat Transfer Systems Laboratory at Purdue University have begun a cooperative research program to investigate the effectiveness of reflux condensation in PWR steam generator U-tubes in the presence of noncondensable gases. The final objectives are to provide local heat transfer data for development of methods to analyze reflux condensation in PWR steam generator U-tubes and to investigate the potential for flooding. Key features of the experimental data reported herein are that they are local data under laminar steam/gas mixture and condensate film flow and they are taken from a test section with dimensions similar to an actual steam generator tube. Steady state data were obtained under various steam and air inlet flow rates and pressures. The data show the significant degrading effect of noncondensable gas on heat transfer coefficients. From the data, correlations for the reflux condensation local heat transfer coefficient and the local Nusselt number under laminar conditions were derived. These experiments are providing essential and unique fundamental data for development of methods to analyze reflux condensation.

  10. Signal processing methods for PWR reactor noise diagnostic system

    International Nuclear Information System (INIS)

    A framework for a PWR reactor noise diagnostic system using various signal processing methods has been investigated. Supposing to treat not only reactor noise data in a stationary linear system but also those in a nonstationary or nonlinear system, the study covers a third-order-correlation of bispectrum, cepstrum analysis, Group Method of Data Handling (GMDH), chaotic quantity, neural network, and wavelet, in addition to Multivariate AutoRegressive analysis and Signal Transmission Path Diagram analysis (MAR/STPD). This paper describes consideration about the methods from viewpoints of theories and applications to PWR reactor noise diagnostic system. The point at the issue in the application system is how to extract many characteristics from the signals whatever states (linear or nonlinear, stationary or nonstationary) may happen in order to get more information and more exact diagnose to support human judgment. From this viewpoint, the paper discusses several signal processing techniques for the PWR diagnostic system. (J.P.N.)

  11. Industry-wide survey of organics in PWR's

    International Nuclear Information System (INIS)

    Interest in organic impurities found in Pressurized Water Reactors (PWR's) has stemmed from several sources. The most serious concern is that organic acids will increase cation conductivity, a parameter that is used to control power plant chemistry. This effect can complicate secondary water monitoring and control. Organics may foul or exhaust makeup demineralizers and condensate polishers, and thus result in increased operating costs or the in leakage of potentially corrosive agents into the steam generators. Some organics, however, such as mopholine and cyclohexylamine may reduce corrosion through oxygen scavenging or surface filming reactions, and may have a positive influence on the pH in areas of local corrosion. At the time this survey began, little information was available on the types or levels of organic impurities that are typically found in PWR's. this survey is intended to provide baseline data for future corrosion testing and to provide fundamental information that will be helpful in refining PWR chemistry guidelines and operating practices

  12. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  13. Numerical simulation of countercurrent gas-liquid flow in a PWR hot leg using VOF method

    International Nuclear Information System (INIS)

    In order to evaluate flow patterns and CCFL (countercurrent flow limitation) characteristics in a PWR hot leg under reflux condensation, numerical simulations have been done using a two-fluid model and a VOF (volume of fluid) method implemented in the CFD software, FLUENT6.3.26. The two-fluid model gave good agreement with CCFL data under low pressure conditions but did not give good results under high pressure steam-water conditions. On the other hand, the VOF method gave good agreement with CCFL data for tests with a rectangular channel but did not give good results for calculations in a circular channel. Therefore, in this paper, the computational grid and schemes were changed in the VOF method, numerical simulations were done for steam-water flows at 1.5 MPa under PWR full-scale conditions with the inner diameter of 0.75 m, and the calculated results were compared with the UPTF data at 1.5 MPa. As a result, the calculated flow pattern was found to be similar to the flow pattern observed in small-scale air-water tests, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa except in the region of a large steam volumetric flux. (author)

  14. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    International Nuclear Information System (INIS)

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system

  15. Evaluation of single liquid primers with organic sulfur compound for bonding between indirect composite material and silver-palladium-copper-gold alloy.

    Science.gov (United States)

    Shimoe, Saiji; Tanoue, Naomi; Satoda, Takahiro; Murayama, Takeshi; Nikawa, Hiroki; Matsumura, Hideo

    2010-01-01

    The purpose of this study was to evaluate the effect of primers on bonding between a silver-palladium-copper-gold alloy and an indirect composite material. Cast disks were air-abraded with alumina, conditioned with one of five primers (Alloy Primer, Luna-Wing Primer, Metal Primer II, Metaltite, M.L. Primer), and bonded with a light-activated indirect composite. Shear bond strengths were determined after 20,000 times of thermocycling. The results showed that four of the primers, except the Luna-Wing Primer, were effective in enhancing the bond strength as compared with the unprimed control group. Of these four primers, Alloy Primer, Metal Primer II, and M.L. Primer exhibited significantly greater bond strengths. It can be concluded that the effectiveness of primers varies considerably according to the organic sulfur compounds added to the solvent, and that care must be taken in selecting priming agents for bonding the composite material and the silver-palladium-copper-gold alloy.

  16. Load-following operation of PWR plants

    International Nuclear Information System (INIS)

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom's N4 plant, KWU's plants, ABB-CE's Systems 80+, and Westinghouse's AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author)

  17. Borssele PWR noise: measurements, analysis and interpretation

    International Nuclear Information System (INIS)

    In the Borssele reactor - a 450 MWe PWR - reactor noise measurements have been performed during four fuel cycles. Measurements were made with a set of ex-core neutron detectors, on one occasion an in-core displacement transducer, and with primary coolant pressure sensors. Digital analysis was applied, where the most powerful tool was the computer programme FAST, which computes auto and cross power spectra for all combinations from a set of many simultaneously recorded signals. Analyses of neutronic signals show a reactivity noise peak at 9.2 Hz, core barrel motion peaks at about 12 and 15 Hz, a damped oscillation at about 2 Hz. Results are given for begin and end of each fuel cycle. The r.m.s. value of the low frequency noise appears to depend linearly on the boron concentration over a wide range. Also some results of primary coolant pressure noise are presented, with coherent peaks below 15 Hz and incoherent peaks above. The second part of the paper describes an alternative way of analyzing and interpreting noise spectra, namely attempts to decompose the neutronic power spectra into physical components, using the information present in the CPSD's of all detector combinations. The components are characterised by their detector position dependency. Effects considered are: uncorrelated noise, global reactivity noise, core motion attenuation noise, and a possible coupling between reactivity and core motion. Results show excellent separation into reactivity and core motion components with possibilities to separate overlapping peaks. Weak peaks become more easily detectable. At low frequencies the decomposition of the spectra is not yet complete, however. (author)

  18. The hold-time effects on the low cycle fatigue behaviors of 316 SS in PWR primary environment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Junho; Hong, Jong-Dae; Seo, Myung-Gyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The effects of the environments on fatigue life of the structural materials used in nuclear power plants (NPPs) were known to be significant according to the extensive test results. Accordingly, the fatigue analysis procedures and the design fatigue curves were proposed in the ASME Code. However, the implication that the existing ASME design fatigue curves did not sufficiently reflect the effect of the operation conditions of nuclear power plants emerged as an issue to be resolved. One of possible reasons to explain the discrepancy is that the laboratory test conditions do not represent the actual plant transients. Therefore, it is necessary to clarify the effects of light water environments on fatigue life while considering more plant-relevant transient conditions such as hold-time. For this reason, this study will focus on the fatigue life of type 316 stainless steel (SS) in the pressurized water reactor (PWR) environments while incorporating the hold-time during the low cycle fatigue (LCF) test in simulated PWR environments. The objective of this study is to characterize the effects of hold-time on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 SS in 310 .deg. C air and simulated PWR environments. To simulate the heat-up and cool-down transient, sub-peak strain holding during the down-hill of strain amplitude was chosen. Currently, LCF tests with 60 seconds holding are in progress. The 0.4, 0.04%/s strain rate condition test results are presented in this study, which shows somewhat longer fatigue life.

  19. The hold-time effects on the low cycle fatigue behaviors of 316 SS in PWR primary environment

    International Nuclear Information System (INIS)

    The effects of the environments on fatigue life of the structural materials used in nuclear power plants (NPPs) were known to be significant according to the extensive test results. Accordingly, the fatigue analysis procedures and the design fatigue curves were proposed in the ASME Code. However, the implication that the existing ASME design fatigue curves did not sufficiently reflect the effect of the operation conditions of nuclear power plants emerged as an issue to be resolved. One of possible reasons to explain the discrepancy is that the laboratory test conditions do not represent the actual plant transients. Therefore, it is necessary to clarify the effects of light water environments on fatigue life while considering more plant-relevant transient conditions such as hold-time. For this reason, this study will focus on the fatigue life of type 316 stainless steel (SS) in the pressurized water reactor (PWR) environments while incorporating the hold-time during the low cycle fatigue (LCF) test in simulated PWR environments. The objective of this study is to characterize the effects of hold-time on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 SS in 310 .deg. C air and simulated PWR environments. To simulate the heat-up and cool-down transient, sub-peak strain holding during the down-hill of strain amplitude was chosen. Currently, LCF tests with 60 seconds holding are in progress. The 0.4, 0.04%/s strain rate condition test results are presented in this study, which shows somewhat longer fatigue life

  20. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  1. The traveller: a new look for PWR fresh fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    Bayley, B.; Stilwell, W.E.; Kent, N.A. [Westinghouse Electric Co., Columbia, SC (United States)

    2004-07-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain.

  2. The traveller: a new look for PWR fresh fuel packages

    International Nuclear Information System (INIS)

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain

  3. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  4. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary

  5. A Practical Primer on Geostatistics

    Science.gov (United States)

    Olea, Ricardo A.

    2009-01-01

    significant methodological implications. HISTORICAL REMARKS As a discipline, geostatistics was firmly established in the 1960s by the French engineer Georges Matheron, who was interested in the appraisal of ore reserves in mining. Geostatistics did not develop overnight. Like other disciplines, it has built on previous results, many of which were formulated with different objectives in various fields. PIONEERS Seminal ideas conceptually related to what today we call geostatistics or spatial statistics are found in the work of several pioneers, including: 1940s: A.N. Kolmogorov in turbulent flow and N. Wiener in stochastic processing; 1950s: D. Krige in mining; 1960s: B. Mathern in forestry and L.S. Gandin in meteorology CALCULATIONS Serious applications of geostatistics require the use of digital computers. Although for most geostatistical techniques rudimentary implementation from scratch is fairly straightforward, coding programs from scratch is recommended only as part of a practice that may help users to gain a better grasp of the formulations. SOFTWARE For professional work, the reader should employ software packages that have been thoroughly tested to handle any sampling scheme, that run as efficiently as possible, and that offer graphic capabilities for the analysis and display of results. This primer employs primarily the package Stanford Geomodeling Software (SGeMS) - recently developed at the Energy Resources Engineering Department at Stanford University - as a way to show how to obtain results practically. This applied side of the primer should not be interpreted as the notes being a manual for the use of SGeMS. The main objective of the primer is to help the reader gain an understanding of the fundamental concepts and tools in geostatistics. ORGANIZATION OF THE PRIMER The chapters of greatest importance are those covering kriging and simulation. All other materials are peripheral and are included for better comprehension of th

  6. A primer on quantum fluids

    CERN Document Server

    Barenghi, Carlo

    2016-01-01

    The aim of this primer is to cover the essential theoretical information, quickly and concisely, in order to enable senior undergraduate and beginning graduate students to tackle projects in topical research areas of quantum fluids, for example, solitons, vortices and collective modes. The selection of the material, both regarding the content and level of presentation, draws on the authors analysis of the success of relevant research projects with newcomers to the field, as well as of the students feedback from many taught and self-study courses on the subject matter. Starting with a brief historical overview, this text covers particle statistics, weakly interacting condensates and their dynamics and finally superfluid helium and quantum turbulence. At the end of each chapter (apart from the first) there will be some exercises. Detailed solutions can be made available to instructors upon request to the authors. .

  7. PrimerMapper: high throughput primer design and graphical assembly for PCR and SNP detection.

    Science.gov (United States)

    O'Halloran, Damien M

    2016-01-01

    Primer design represents a widely employed gambit in diverse molecular applications including PCR, sequencing, and probe hybridization. Variations of PCR, including primer walking, allele-specific PCR, and nested PCR provide specialized validation and detection protocols for molecular analyses that often require screening large numbers of DNA fragments. In these cases, automated sequence retrieval and processing become important features, and furthermore, a graphic that provides the user with a visual guide to the distribution of designed primers across targets is most helpful in quickly ascertaining primer coverage. To this end, I describe here, PrimerMapper, which provides a comprehensive graphical user interface that designs robust primers from any number of inputted sequences while providing the user with both, graphical maps of primer distribution for each inputted sequence, and also a global assembled map of all inputted sequences with designed primers. PrimerMapper also enables the visualization of graphical maps within a browser and allows the user to draw new primers directly onto the webpage. Other features of PrimerMapper include allele-specific design features for SNP genotyping, a remote BLAST window to NCBI databases, and remote sequence retrieval from GenBank and dbSNP. PrimerMapper is hosted at GitHub and freely available without restriction. PMID:26853558

  8. PrimerMapper: high throughput primer design and graphical assembly for PCR and SNP detection.

    Science.gov (United States)

    O'Halloran, Damien M

    2016-01-01

    Primer design represents a widely employed gambit in diverse molecular applications including PCR, sequencing, and probe hybridization. Variations of PCR, including primer walking, allele-specific PCR, and nested PCR provide specialized validation and detection protocols for molecular analyses that often require screening large numbers of DNA fragments. In these cases, automated sequence retrieval and processing become important features, and furthermore, a graphic that provides the user with a visual guide to the distribution of designed primers across targets is most helpful in quickly ascertaining primer coverage. To this end, I describe here, PrimerMapper, which provides a comprehensive graphical user interface that designs robust primers from any number of inputted sequences while providing the user with both, graphical maps of primer distribution for each inputted sequence, and also a global assembled map of all inputted sequences with designed primers. PrimerMapper also enables the visualization of graphical maps within a browser and allows the user to draw new primers directly onto the webpage. Other features of PrimerMapper include allele-specific design features for SNP genotyping, a remote BLAST window to NCBI databases, and remote sequence retrieval from GenBank and dbSNP. PrimerMapper is hosted at GitHub and freely available without restriction.

  9. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  10. Electrostatic Discharge testing of propellants and primers

    Energy Technology Data Exchange (ETDEWEB)

    Berry, R.B.

    1994-02-01

    This report presents the results of testing of selected propellants and primers to Electrostatic Discharge (ESD) characteristic of the human body. It describes the tests and the fixturing built to accommodate loose material (propellants) and the packed energetic material of the primer. The results indicate that all powders passed and some primers, especially the electric primers, failed to pass established requirements which delineate insensitive energetic components. This report details the testing of components and materials to four ESD environments (Standard ESD, Severe ESD, Modified Standard ESD, and Modified Severe ESD). The purpose of this study was to collect data based on the customer requirements as defined in the Sandia Environmental Safety & Health (ES&H) Manual, Chapter 9, and to define static sensitive and insensitive propellants and primers.

  11. Studies of a small PWR for onsite industrial power

    International Nuclear Information System (INIS)

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application

  12. Studies of a small PWR for onsite industrial power

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, O.H.; Smith, W.R.

    1977-04-19

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application.

  13. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author)

  14. Experience and reliability of Framatome ANP's PWR and BWR fuel

    International Nuclear Information System (INIS)

    Based on three decades of fuel supply to 169 PWR and BWR plants on four continents, Framatome ANP has a very large database from operating experience feedback. The performance of Framatome PWR and BWR fuel is discussed for the period 1992-2001 with special emphasis on fuel failures, countermeasures and their effectiveness. While PWR fuel performance in most reactors has been good, the performance in some years did suffer from special circumstances that caused grid-to-rod fretting failures in few PWRs. After solving this problem, fuel of all types showed high reliability again. Especially the current PWR fuel products AFA 3G, HTP, Mark B and Mark BW showed a very good operating performance. Fuel reliability of Framatome ANP BWR fuel has been excellent over the last decade with average annual fuel rod failure rates under 1x10-5 since 1991. More than 40% of all BWR fuel failures in the 1992-2001 decade were caused by debris fretting. The debris problem has been remedied with the FUELGUARDTM lower tie plate, and by reactor operators' efforts to control the sources of debris. PCI, the main failure mechanism in former periods, affected only 10 rods. All of these rods had non-liner cladding. (author)

  15. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  16. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    The EDF experience acquired permits to allow the PWR fuel performances and to make use of better management. In this domain low progress can be given considerable financial profits. The industrial and commercial structures, the time constant of the fuel cycle, has for consequence that the electric utilities can take advantage only progressively of the expected profits

  17. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  18. Long term Integrity of PWR Spent Fuel in Dry Storage

    International Nuclear Information System (INIS)

    The newly established organization KRMC (Korea radioactive waste management corporation) which is responsible for all kinds of radioactive waste generated in the Republic of Korea launched the PWR spent fuel dry storage research project in June 2009. This project has objectives to develop a storage system and evaluate the integrity of PWR fuel in dry storage. The project consists of three steps. At first step, it would develop own degradation models by referring to pre-exist good models and develop the hot test scenarios. Second step, test facilities would be constructed and used for testing the degradation behaviour in each mechanisms and in total. As a final step, total evaluation code would be developed by integrating each degradation model produced in the first step and the test data produced in the second step. All the activities would be summarized into a report and applied to licensing work. The Republic of Korea PWR spent fuels have unique characteristics of various fuel types (array type, clad material) and high capacity factor (maximum usage of fuel which is bad for integrity). These facts could impact on the research ranges of experimental data needed for degradation evaluation. In this research, spent fuel performance data concerning long term dry storage will be analysed and the major degradation mechanisms like creep and hydride behaviour will be studied and proposed for Korean PWR spent fuels

  19. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    A study about fuel management in PWR reactors, where not only the conventional uranium cycle is considered, but also the thorium cycle as an alternative is presented. The final results are presented in terms of U3O8 demand and SWU and the approximate costs of the principal stages of the fuel cycle, comparing with the stardand cycle without recycling. (E.G.)

  20. SBE primer : multiplexing minisequencing-based genotyping

    Energy Technology Data Exchange (ETDEWEB)

    Kaderali, L. (Lars); Deshpande, A. (Alina); Uribe-Romeo, F. J. (Francisco J.); Schliep, A.; Torney, D. C. (David C.)

    2002-01-01

    Single-nucleotide polymorphism (SNP) analysis is a powerful tool for mapping and diagnosing disease-related alleles. Most of the known genetic diseases are caused by point mutations, and a growing number of SNPs will be routinely analyzed to diagnose genetic disorders. Mutation analysis by polymerase mediated single-base primer extension (minisequencing) can be massively parallelized using for example DNA microchips or flow cytometry with microspheres as solid support. By adding a unique oligonucleotide tag to the 5-inch end of the minisequencing primer and attaching the complementary anti-tag to the array or bead surface, the assay can be 'demultiplexed'. However, such high-throughput scoring of SNPs requires a high level of primer multiplexing in order to analyze multiple loci in one assay, thus enabling inexpensive and fast polymorphism scoring. Primers can be chosen from either the plus or the minus strand, and primers used in the same experiment must not bind to one another. To genotype a given number of polymorphic sites, the question is which primer to use for each SNP, and which primers to group into the same experiment. Furthermore, a crosshybridization-free tag/anti-tag code is required in order to sort the extended primers to the corresponding microspheres or chip spots. These problems pose challenging algorithmic questions. We present a computer program lo automate the design process for the assay. Oligonucleotide primers for the reaction are automatically selected by the software, a unique DNA tag/anti-tag system is generated, and the pairing of primers and DNA-Tags is automatically done in a way to avoid any crossreactivity. We report first results on a 45-plex genotyping assay, indicating that minisequencing can be adapted to be a powerful tool for high-throughput, massively parallel genotyping.

  1. PHUSER (Primer Help for USER): a novel tool for USER fusion primer design

    DEFF Research Database (Denmark)

    Olsen, Lars Rønn; Hansen, Niels Bjørn; Bonde, Mads;

    2011-01-01

    , a novel tool for designing primers specifically for USER fusion and USER cloning applications. We also present proof-of-concept experimental validation of its functionality. PHUSER offers quick and easy design of PCR optimized primers ensuring directionally correct fusion of fragments into a plasmid......Uracil-Specific Exision Reagent (USER) fusion is a recently developed technique that allows for assembly of multiple DNA fragments in a few simple steps. However, designing primers for USER fusion is both tedious and time consuming. Here, we present the Primer Help for USER (PHUSER) software...... containing a customizable USER cassette. Designing primers using PHUSER ensures that the primers have similar annealing temperature (Tm), which is essential for efficient PCR. PHUSER also avoids identical overhangs, thereby ensuring correct order of assembly of DNA fragments. All possible primers...

  2. Single PWR spent fuel assembly heat transfer data for computer code evaluations

    International Nuclear Information System (INIS)

    The descriptions and results of two separate heat transfer tests designed to investigate the dry storage of commercial PWR spent fuel assemblies are presented. Presented first are descriptions and selected results from the Fuel Temperature Test performed at the Engine Maintenance and Disassembly facility on the Nevada Test Site. An actual spent fuel assembly from the Turkey Point Unit Number 3 Reactor with a decay heat level of 1.17 KW, was installed vertically in a test stand mounted canister/liner assembly. The boundary temperatures were controlled and the canister backfill gases were alternated between air, helium and vacuum to investigate the primary heat transfer mechanisms of convection, conduction and radiation. The assembly temperature profiles were experimentally measured using installed thermocouple instrumentation. Also presented are the results from the Single Assembly Heat Transfer Test designed and fabricated by Allied General Nuclear Services, under contract to the Department of Energy, and ultimately conducted by the Pacific Northwest Laboratory. For this test, an electrically heated 15 x 15 rod assembly was used to model a single PWR spent fuel assembly. The electrically heated model fuel assembly permitted various ''decay heat'', levels to be tested; 1.0 KW and 0.5 KW were used for these tests. The model fuel assembly was positioned within a prototypic fuel tube and in turn placed within a double-walled sealed cask. The complete test assembly could be positioned at any desired orientation (horizontal, vertical, and 250 from horizontal for the present work) and backfilled as desired (air, helium, or vacuum). Tests were run for all combinations of ''decay heat,'' backfill, and orientation. Boundary conditions were imposed by temperature controlled guard heaters installed on the cask exterior surface

  3. VOF Calculations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    Directory of Open Access Journals (Sweden)

    M. Murase

    2012-01-01

    Full Text Available We improved the computational grid and schemes in the VOF (volume of fluid method with the standard − turbulent model in our previous study to evaluate CCFL (countercurrent flow limitation characteristics in a full-scale PWR hot leg (750 mm diameter, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa. In this paper, therefore, to evaluate applicability of the VOF method to different fluid properties and a different scale, we did numerical simulations for full-scale air-water conditions and the 1/15-scale air-water tests (50 mm diameter, respectively. The results calculated for full-scale conditions agreed well with CCFL data and showed that CCFL characteristics in the Wallis diagram were mitigated under 1.5 MPa steam-water conditions comparing with air-water flows. However, the results calculated for the 1/15-scale air-water tests greatly underestimated the falling water flow rates in calculations with the standard − turbulent model, but agreed well with the CCFL data in calculations with a laminar flow model. This indicated that suitable calculation models and conditions should be selected to get good agreement with data for each scale.

  4. Corrosion fatigue initiation behaviour of wrought austenitic stainless pipe steels under simulated BWR/HWC and PWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Leber, H.J.; Ritter, S.; Seifert, H.P [Paul Scherrer Institute, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen PSI (Switzerland)

    2011-07-01

    The corrosion fatigue (CF) initiation and short crack growth behavior of different low-carbon and stabilized austenitic stainless steels was characterized under simulated BWR and primary PWR conditions by cyclic fatigue tests with sharply notched fracture mechanics specimens in the temperature range from 70 to 320 C. Environmental reduction of fatigue initiation life was observed in all stainless steels at strain rates {<=} 0.1 %/s in BWR and PWR environment. The stationary short crack CF crack growth rates after crack advances of 50 to 300 {mu}m from the notch-root were in the typical range of corresponding results from tests with long cracks (pre-cracked specimens) and also showed the same system parameter response. The effect of environment on the initiation process ({Delta}a = 10 {mu}m) was relevantly stronger than on the subsequent stationary short crack growth. Both, under BWR/HWC and PWR conditions, a relevant environmental reduction of fatigue initiation life occurred for the combination of temperatures {>=} 100 C, notch strain rates {<=} 0.1 %/s and notch strain amplitudes {>=} 0.3 %. If these conjoint threshold conditions were simultaneously satisfied, the environmental enhancement increased with decreasing strain rate and increasing temperature. Material and water chemistry parameters usually only had a little effect. Sensitization affected the CF behavior under highly oxidizing BWR/NWC conditions only. Preliminary block loading experiments did not reveal significant static load hold period effects on the technical corrosion fatigue initiation life. If the critical requirements were satisfied, the BWR/HWC and PWR environments usually resulted in acceleration of short fatigue crack growth by a factor of 5 to 20 with respect to air. Solution annealed steels showed slightly shorter CF initiation lives, but also lower stationary short CF crack growth rates under BWR/HWC and PWR conditions with low ECPs than under highly oxidizing BWR/NWC conditions. A very

  5. Primer on spontaneous heating and pyrophoricity

    Energy Technology Data Exchange (ETDEWEB)

    1994-12-01

    This primer was prepared as an information resource for personnel responsible for operation of DOE nuclear facilities. It has sections on combustion principles, spontaneous heating/ignition of hydrocarbons and organics, pyrophoric gases and liquids, pyrophoric nonmetallic solids, pyrophoric metals (including Pu and U), and accident case studies. Although the information in this primer is not all-encompassing, it should provide the reader with a fundamental knowledge level sufficient to recognize most spontaneous combustion hazards and how to prevent ignition and widespread fires. This primer is provided as an information resource only, and is not intended to replace any fire protection or hazardous material training.

  6. Primers-4-Yeast: a comprehensive web tool for planning primers for Saccharomyces cerevisiae.

    Science.gov (United States)

    Yofe, Ido; Schuldiner, Maya

    2014-02-01

    The budding yeast Saccharomyces cerevisiae is a key model organism of functional genomics, due to its ease and speed of genetic manipulations. In fact, in this yeast, the requirement for homologous sequences for recombination purposes is so small that 40 base pairs (bp) are sufficient. Hence, an enormous variety of genetic manipulations can be performed by simply planning primers with the correct homology, using a defined set of transformation plasmids. Although designing primers for yeast transformations and for the verification of their correct insertion is a common task in all yeast laboratories, primer planning is usually done manually and a tool that would enable easy, automated primer planning for the yeast research community is still lacking. Here we introduce Primers-4-Yeast, a web tool that allows primers to be designed in batches for S. cerevisiae gene-targeting transformations, and for the validation of correct insertions. This novel tool enables fast, automated, accurate primer planning for large sets of genes, introduces consistency in primer planning and is therefore suggested to serve as a standard in yeast research. Primers-4-Yeast is available at: http://www.weizmann.ac.il/Primers-4-Yeast

  7. Porous Media Primer for Physicists

    Science.gov (United States)

    Hunt, Allen; Ewing, Robert

    The study of soils and rocks is the province of many different disciplines. These disciplines have historically focused on applications rather than understanding, and this pragmatic approach has led to some cutting of corners. Furthermore, the different disciplines have different goals, so they have developed their own peculiar vocabulary, insights, and biases. For example, petroleum engineers generally work with consolidated rock, so the concept of a particle size distribution is not as central to their thinking as it is to a soil scientist. Meanwhile, soil scientists working with just two fluids - air and water - can frequently get away with assuming that air is infinitely compressible (and has density and viscosity of zero); petroleum engineers working with multiple flowing gases and liquids must consider all fluid phases in concert. Insofar as the structure of the medium is concerned, the material presented here tends to be centered on soil physics, but we have attempted to make contact with other disciplines in important cases.

  8. Multiplexing Short Primers for Viral Family PCR

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, S N; Hiddessen, A L; Hara, C A; Williams, P L; Wagner, M; Colston, B W

    2008-06-26

    We describe a Multiplex Primer Prediction (MPP) algorithm to build multiplex compatible primer sets for large, diverse, and unalignable sets of target sequences. The MPP algorithm is scalable to larger target sets than other available software, and it does not require a multiple sequence alignment. We applied it to questions in viral detection, and demonstrated that there are no universally conserved priming sequences among viruses and that it could require an unfeasibly large number of primers ({approx}3700 18-mers or {approx}2000 10-mers) to generate amplicons from all sequenced viruses. We then designed primer sets separately for each viral family, and for several diverse species such as foot-and-mouth disease virus, hemagglutinin and neuraminidase segments of influenza A virus, Norwalk virus, and HIV-1.

  9. DNA sequencing by synthesis with degenerate primers

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    The degenerate primer-based sequencing Was developed by a synthesis method(DP-SBS)for high-throughput DNA sequencing,in which a set of degenerate primers are hybridized on the arrayed DNA templates and extended by DNA polymerase on microarrays.In this method,adifferent set of degenerate primers containing a give nnumber(n)of degenerate nucleotides at the 3'-ends were annealed to the sequenced templates that were immobilized on the solid surface.The nucleotides(n+1)on the template sequences were determined by detecting the incorporation of fluorescent labeled nucleotides.The fluorescent labeled nucleotide was incorporated into the primer in a base-specific manner after the enzymatic primer extension reactions and nine-base length were read out accurately.The main advanmge of the DP-SBS is that the method only uses very conventional biochemical reagents and avoids the complicated special chemical reagents for removing the labeled nucleotides and reactivating the primer for further extension.From the present study,it is found that the DP-SBS method is reliable,simple,and cost-effective for laboratory-sequencing a large amount of short DNA fragments.

  10. UniPrimer: A Web-Based Primer Design Tool for Comparative Analyses of Primate Genomes

    OpenAIRE

    Nomin Batnyam; Jimin Lee; Jungnam Lee; Seung Bok Hong; Sejong Oh; Kyudong Han

    2012-01-01

    Whole genome sequences of various primates have been released due to advanced DNA-sequencing technology. A combination of computational data mining and the polymerase chain reaction (PCR) assay to validate the data is an excellent method for conducting comparative genomics. Thus, designing primers for PCR is an essential procedure for a comparative analysis of primate genomes. Here, we developed and introduced UniPrimer for use in those studies. UniPrimer is a web-based tool that designs PCR-...

  11. Adhesion and thermal stability enhancement of IZO films by adding a primer layer on polycarbonate substrate

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xuan; Zhang, Xiaofeng; Yan, Yue; Zhong, Yanli; Li, Lei; Zhang, Guanli [Beijing Institute of Aeronautical Materials (BIAM), Haidian District, Beijing, 100095 (China)

    2015-04-01

    A silicone-based primer layer was developed to improve the adhesion and thermal stability of amorphous transparent indium zinc oxide (IZO) films on polycarbonate (PC). The IZO films deposited by direct current magnetron sputtering at room temperature on primer-treated and untreated PCs were evaluated ex situ in terms of surface morphology, adhesion, optical, and electrical properties during annealing at 120 C in air. Nano-scratch tests indicated the adhesion of IZO films on primer-treated substrates was superior to that on untreated PCs. This superior adhesion can be attributed to the strong Si-O-Si inorganic bonds abundant in the primer layer and better matches of the primer layer in the terms of thermal expansion to the IZO. Moreover, the electrical resistivity of IZO films prepared on primer-treated PCs remained stable during the annealing treatment, whereas those of IZO films on untreated PCs presented a continuously increasing trend, which was attributed to the decrease in carrier concentration that resulted from oxygen adsorption. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. Report on the PWR-radiation protection/ALARA Committee

    International Nuclear Information System (INIS)

    In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and information relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance

  13. Industrial assessment of nonbackfittable PWR design modifications. Final report

    International Nuclear Information System (INIS)

    As part of the US Department of Energy's Advanced Reactor Design Study, various nonbackfittable PWR design modifications were evaluated to determine their potential for improved uranium utilization and commercial viability. Combustion Engineering, Inc. contributed to this effort through participation in the Battelle Pacific Northwest Laboratory industrial assessment of such design modifications. Seven modifications, including the use of higher primary system temperatures and pressures, rapid-frequent refueling, end-of-cycle stretchout, core periphery modifications, radial blankets, low power density cores, and small PWR assemblies, were evaluated with respect to uranium utilization, economics, technical and operational complexity, and several other subjective considerations. Rapid-frequent refueling was judged to have the highest potential although it would probably not be economical for the majority of reactors with the design assumptions used in this assessment

  14. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Flow clogging characteristics were investigated based on data for the relation of pressure loss and flow velocity during flow clogging due to debris accumulation. Deposition of chemical precipitates on the fuel cladding using an electrically heated rod was investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis with a thermal-hydraulic code on the downstream effect has shown that the core could be cooled because the core inlet flow compensates a evaporation of coolant due to the decay-heat even if core inlet was 99% clogged just after the ECCS recirculation operation started during the cold-leg break LOCA in PWR plants. (author)

  15. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the chemical effect and the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Chemical effect tests show that corrosion of carbon steel and galvanized steal may come to be important in domestic plants, in addition to corrosion of aluminum and insulator which has been considered dominant in the chemical effect. With respect to the downstream effect, deposition of chemical precipitates on the fuel cladding using an electrically heated rod is investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis on the downstream effect has shown that even if core inlet was completely clogged just after the recirculation operation started during LOCA in PWR plants, although upper part of core may be uncovered temporary and cladding temperature increased, core could be cooled by coolant injection through the hot-leg. (author)

  16. A concept of PWR using plate and shell heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  17. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  18. A concept of PWR using plate and shell heat exchangers

    International Nuclear Information System (INIS)

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  19. On catholyte application for hydrogen water chemistry in PWR

    International Nuclear Information System (INIS)

    Considering liquid water as a chemical compound with a wide band gap shows that its Redox potential as Fermi level in the band gap is the measurable characteristic of a non-stoichiometric aqueous coolant in recirculation system of PWR. The hypo-stoichiometric state with the negative Redox potential is realized when Fermi level is shifted to the bottom of conduction band. This state can be fixed by the electro-reduced water (catholyte) of the alkaline solution. Then, the hydride anions (H3O-) as proton acceptors and the hydrox-onium radicals (H3O) as electron donors are emerged in the alkaline catholyte and form hydrated clusters (AH)n(H2O)m of alkaline hydride. These particles as very strong reducers have a molar portion more than the gaseous hydrogen in the aqueous coolant and are the effective remedy for holding the negative Redox potential as an effect of hydrogen water chemistry in PWR. (authors)

  20. A burnup credit calculation methodology for PWR spent fuel transportation

    International Nuclear Information System (INIS)

    A burnup credit calculation methodology for PWR spent fuel transportation has been developed and validated in CEA/Saclay. To perform the calculation, the spent fuel composition are first determined by the PEPIN-2 depletion analysis. Secondly the most important actinides and fission product poisons are automatically selected in PEPIN-2 according to the reactivity worth and the burnup for critically consideration. Then the 3D Monte Carlo critically code TRIMARAN-2 is used to examine the subcriticality. All the resonance self-shielded cross sections used in this calculation system are prepared with the APOLLO-2 lattice cell code. The burnup credit calculation methodology and related PWR spent fuel transportation benchmark results are reported and discussed. (authors)

  1. Three basic options for the management of PWR waste

    International Nuclear Information System (INIS)

    Relying on the national practices of France, Germany and Belgium, three reference management routes for PWR wastes were drawn up and subsequently evaluated in terms of costs and radiological impact. It was thus demonstrated that safety regulations and technical redundancies, especially for off-gas treatment, liquid waste processing and dry solid waste treatment, play an important part in the cost associated with each route. The analysis of the different treatment options for mixed solid low level waste highlighted the low cost effectiveness of incineration as compared to compaction. Whatever the scenario investigated, the disposal costs of PWR wastes proved to be quite marginal in the overall cost. The radiological impact associated with each route was assessed through individual doses resulting from liquid and gaseous effluents. This theoretical exercise included some sensitivity studies performed on a selection of important parameters

  2. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

  3. Control room dose analysis for Maanshan PWR plant during design basis loss of coolant accident

    International Nuclear Information System (INIS)

    To address the issue identified in USNRC's Generic Letter 2003-1 that the unfiltered air in-leakage rate through plant's control room during design basis accident may exceed that assumed in the licensing analysis and thus threat the control room habitability, the control room radiation dose analysis of Maanshan PWR plant has to be re-performed to determine the allowable unfiltered air in-leakage rate. The allowable unfiltered air in-leakage rate is to be determined in such a way that the calculated whole body dose in the control room during the most limiting design basis accident must meet the criteria set forth in 10 CFR 50 Appendix A General Design Criteria (GDC) 19. The determined allowable air in-leakage rate is then employed as an acceptable limit to be met by the control room in-leakage test. In this study, the Maanshan plant control room dose analysis model during loss of coolant accident (LOCA) has been established based on USNRC's RADTRAD computer code. Different release and transport paths have been incorporated in this model, including containment leakage, engineered safety feature (ESF) leakage, and control room filtered and un-filtered air in-leakage. The RADTRAD calculation results are compared with Final Safety Analysis Report (FSAR) results to assure that overall consistency is reached. Finally, considering the uncertainties and margin to be maintained between RADTRAD calculation results and GDC-19 dose limits, an allowable unfiltered air in-leakage rate for control room habitability application during LOCA has been well defined. (author)

  4. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  5. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    The key computer programs currently used by the analysis procedure of the ORNL-PWR Blowdown Heat Transfer Separate Effects Program are overviewed with particular emphasis placed on their interrelationships. The major modeling and calculational programs, COBRA, ORINC, ORTCAL, PINSIM, and various versions of RELAP4, are summarized and placed into the perspective of the procedure. The supportive programs, REDPLT, ORCPLT, BDHTPLOT, OXREPT, and OTOCI, and their uses are described

  6. Improvement on main control room for Japanese PWR plants

    International Nuclear Information System (INIS)

    The main control room which is the information center of nuclear power plant has been continuously improved utilizing the state of the art ergonomics, a high performance computer, computer graphic technologies, etc. For the latest Japanese Pressurized Water Reactor (PWR) plant, the CRT monitoring system is applied as the major information source for facilitating operators' plant monitoring tasks. For an operating plant, enhancement of monitoring and logging functions has been made adopting a high performance computer

  7. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.)

  8. Propagation of nuclear data Uncertainties for PWR core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cabellos, O.; Castro, E.; Ahnert, C.; Holgado, C. [Dept. of Nuclear Engineering, Universidad Politecnica de Madrid, Madrid (Spain)

    2014-06-15

    An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for {sup 235,238}U, {sup 239}Pu, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

  9. In-pile test of Qinshan PWR fuel bundle

    International Nuclear Information System (INIS)

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  10. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Cho, Dong Keun; Chun, Kwan Sik; Lee, Jong Youl; Kim, Seong Ki; Kim, Seong Soo; Lee, Yang

    2005-11-15

    This report outlined the results of the safety assessment of the canisters for the PWR spent fuels which will be used in the KRS. All safety analyses including criticality and radiation shielding analyses, mechanical analyses, thermal analyses, and containment analyses were performed. The reference PWR spent fuels were in the 17x17 and determined to have 45,000 MWD/MTU burnup. The canister consists of copper outer shell and nodular cast iron inner structure with diameter of 102 cm and height of 483 cm. Criticality safety was checked for normal and abnormal conditions. It was assumed that the integrity of engineered barriers is preserved and saturated with water of 1.0g/cc for normal condition. For the abnormal condition container and bentonite was assumed to disappear, which allows the spent fuel to be surrounded by water with the most reactive condition. In radiation shielding analysis it was investigated that the absorbed dose at the surface of the canister met the safety limit. The structural analysis was conducted considering three load conditions, normal, extreme, and rock movement condition. Thermal analysis was carried out for the case that the canister with four PWR assemblies was deposited in the repository 500 meter below the surface with 40 m tunnel spacing and 6 m deposition hole spacing. The results of the safety assessment showed that the proposed KDC-1 canister met all the safety limits.

  11. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  12. Validation of gadolinium burnout using PWR benchmark specification

    International Nuclear Information System (INIS)

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor Kinf, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157

  13. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  14. Actinides transmutation - a comparison of results for PWR benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  15. QFLOOD-GT: a program for predicting PWR reflood

    International Nuclear Information System (INIS)

    A description is given of the present version of the QFLOOD-GT program for predicting the reflood stage of a large-break PWR loss-of-coolant accident. QFLOOD-GT has been developed from an earlier forced-reflood program which, using a conduction-controlled model for rewetting speed, gave good agreement with the FLECHT SEASET experiments. This earlier program has been incorporated into QFLOOD-GT as a subroutine called QFLOOD; in addition a downcomer model has been included in order to allow calculation of gravity reflood, and a computational scheme has been devised to simulate the chimney effect (the unequal distribution of inlet flow between hot and cool regions of the core). No quantitative comparisons between QFLOOD-GT predictions and integral-test data have yet been carried out, so the modelling decisions implemented in the program are at this stage unvalidated. Preliminary testing of the program has produced results which are for the most part qualitatively satisfactory. Calculations for indicative PWR conditions suggest that the chimney effect has a significant beneficial effect during PWR reflood, a conclusion in accordance with the findings of the Japanese 2D/3D experiments. (author)

  16. PWR experimental benchmark analysis using WIMSD and PRIDE codes

    International Nuclear Information System (INIS)

    Highlights: • PWR experimental benchmark calculations were performed using WIMSD and PRIDE codes. • Various models for lattice cell homogenization were used. • Multiplication factors, power distribution and reaction rates were studied. • The effect of cross section libraries on these parameters was analyzed. • The results were compared with experimental and reported results. - Abstract: The PWR experimental benchmark problem defined by ANS was analyzed using WIMSD and PRIDE codes. Different modeling methodologies were used to calculate the infinite and effective multiplication factors. Relative pin power distributions were calculated for infinite lattice and critical core configurations, while reaction ratios were calculated for infinite lattice only. The discrete ordinate method (DSN) and collision probability method (PERSEUS) were used in each calculation. Different WIMSD cross-section libraries based on ENDF/B-VI.8, ENDF/B-VII.0, IAEA, JEF-2.2, JEFF-3.1 and JENDL-3.2 nuclear data files were also employed in the analyses. Comparison was made with experimental data and other reported results in order to find a suitable strategy for PWR analysis

  17. PWR primary coolant sample lines - problems with measurement of corrosion products and experimental proposals for Ringhals PWR

    International Nuclear Information System (INIS)

    Coolant samples are drawn from PWR primary circuits through long narrow tubes. Concern that interaction with the sample line walls (by deposition and release) can result in inaccurate measurement of corrosion product concentrations has recently intensified after several observations of a dependence on sample line flow rate. Particularly significant instances of this have been observed at Ringhals PWR. A further problem is that measured concentrations show spurious transient increases after valving in the sample line. Sampling behaviour is complex since it involves particulate as well as soluble material, and deposition and release as well as localised phenomena associated with crud traps within the sample line. The present report has threefold function, firstly to review instances of anomalous sample line behaviour and secondly to present a basic theoretical background to aid interpretation of such behaviour. The third and most important function is to suggest plant measurements which might be made at Ringhals PWR to understand better the response of the sampling system by quantifying the effects due to corrosion product deposition on, and release from, sample line walls. (author)

  18. Climate Change, Health, and Communication: A Primer.

    Science.gov (United States)

    Chadwick, Amy E

    2016-01-01

    Climate change is one of the most serious and pervasive challenges facing us today. Our changing climate has implications not only for the ecosystems upon which we depend, but also for human health. Health communication scholars are well-positioned to aid in the mitigation of and response to climate change and its health effects. To help theorists, researchers, and practitioners engage in these efforts, this primer explains relevant issues and vocabulary associated with climate change and its impacts on health. First, this primer provides an overview of climate change, its causes and consequences, and its impacts on health. Then, the primer describes ways to decrease impacts and identifies roles for health communication scholars in efforts to address climate change and its health effects.

  19. Climate Change, Health, and Communication: A Primer.

    Science.gov (United States)

    Chadwick, Amy E

    2016-01-01

    Climate change is one of the most serious and pervasive challenges facing us today. Our changing climate has implications not only for the ecosystems upon which we depend, but also for human health. Health communication scholars are well-positioned to aid in the mitigation of and response to climate change and its health effects. To help theorists, researchers, and practitioners engage in these efforts, this primer explains relevant issues and vocabulary associated with climate change and its impacts on health. First, this primer provides an overview of climate change, its causes and consequences, and its impacts on health. Then, the primer describes ways to decrease impacts and identifies roles for health communication scholars in efforts to address climate change and its health effects. PMID:26580230

  20. Microsatellite primers for fungus-growing ants

    DEFF Research Database (Denmark)

    Villesen, Palle; Gertsch, P J; Boomsma, JJ

    2002-01-01

    We isolated five polymorphic microsatellite loci from a library of two thousand recombinant clones of two fungus-growing ant species, Cyphomyrmex longiscapus and Trachymyrmex cf. zeteki. Amplification and heterozygosity were tested in five species of higher attine ants using both the newly...... developed primers and earlier published primers that were developed for fungus-growing ants. A total of 20 variable microsatellite loci, developed for six different species of fungus-growing ants, are now available for studying the population genetics and colony kin-structure of these ants....

  1. Microsatellite Primers for Fungus-Growing Ants

    DEFF Research Database (Denmark)

    Villesen Fredsted, Palle; Gertsch, Pia J.; Boomsma, Jacobus Jan (Koos)

    2002-01-01

    We isolated five polymorphic microsatellite loci from a library of two thousand recombinant clones of two fungus-growing ant species, Cyphomyrmex longiscapus and Trachymyrmex cf. zeteki. Amplification and heterozygosity were tested in five species of higher attine ants using both the newly...... developed primers and earlier published primers that were developed for fungus-growing ants. A total of 20 variable microsatellite loci, developed for six different species of fungus-growing ants, are now available for studying the population genetics and colony kin-structure of these ants....

  2. Development of Specific Primer for Tricholoma matsutake

    OpenAIRE

    Kim, Jang-Han; Han, Yeong-Hwan

    2009-01-01

    In this study, in an effort to develop a method for the molecular detection of Tricholoma matsutake in Korea from other closely related Tricholomataceae, a species-specific PCR primer pair, TmF and TmR, was designed using nuclear ribosomal intertranscribed spacer (ITS) sequences. The DTmF and DTmR sequences were 5'-CCTGACGCCAATCTTTTCA-3' and 5'-GGAGAGCAGACTTGTGAGCA-3', respectively. The PCR primers reliably amplified only the ITS sequences of T. matsutake, and not those of other species used ...

  3. EPRI PWR Safety and Relief Valve Test Program: test condition justification report

    Energy Technology Data Exchange (ETDEWEB)

    Hosler, J.

    1982-12-01

    In response to NUREG 0737, Item II.D.1.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented.

  4. Shear Bond Strength of MDP-Containing Self-Adhesive Resin Cement and Y-TZP Ceramics: Effect of Phosphate Monomer-Containing Primers

    Directory of Open Access Journals (Sweden)

    Jin-Soo Ahn

    2015-01-01

    Full Text Available Purpose. This study was conducted to evaluate the effects of different phosphate monomer-containing primers on the shear bond strength between yttria-tetragonal zirconia polycrystal (Y-TZP ceramics and MDP-containing self-adhesive resin cement. Materials and Methods. Y-TZP ceramic surfaces were ground flat with #600-grit SiC paper and divided into six groups (n=10. They were treated as follows: untreated (control, Metal/Zirconia Primer, Z-PRIME Plus, air abrasion, Metal/Zirconia Primer with air abrasion, and Z-PRIME Plus with air abrasion. MDP-containing self-adhesive resin cement was applied to the surface-treated Y-TZP specimens. After thermocycling, a shear bond strength test was performed. The surfaces of the Y-TZP specimens were analyzed under a scanning electron microscope. The bond strength values were statistically analyzed using one-way analysis of variance and the Student–Newman–Keuls multiple comparison test (P<0.05. Results. The Z-PRIME Plus treatment combined with air abrasion produced the highest bond strength, followed by Z-PRIME Plus application, Metal/Zirconia Primer combined with air abrasion, air abrasion alone, and, lastly, Metal/Zirconia Primer application. The control group yielded the lowest results (P<0.05. Conclusion. The application of MDP-containing primer resulted in increased bond strength between Y-TZP ceramics and MDP-containing self-adhesive resin cements.

  5. De neurocirujano a primer ministro de Salud de la Argentina

    Directory of Open Access Journals (Sweden)

    Karina Inés Ramacciotti

    2008-01-01

    Full Text Available La trayectoria y los vínculos que entabló Ramón Carrillo con anterioridad a ejercer el cargo de primer secretario de Salud Pública en la Argentina (1946 no han sido objeto de estudio pormenorizado. Así pues en este artículo se analizarán en primer lugar, una serie de cartas de lectores publicadas en La Semana Médica en los primeros años de la década del '40 del siglo XX. Estas notas permiten comprender las disputas internas que se produjeron en la Facultad de Ciencias Médicas al producirse el concurso de Titular de Neurocirugía de la Universidad de Buenos Aires. En segundo lugar, se revisará cómo Carrillo pasa de ocupar este prestigioso cargo académico a convertirse en decano interi- no de la Facultad de Ciencias Médicas. Son las relaciones que anuda durante estos años las que lo posicionan en un escenario político privilegiado para alcanzar un relevante puesto en la administración pública.

  6. PWR safety and relief valve test program. Valve selection/juftification report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    NUREG 0578 required that full-scale testing be performed on pressurizer safety valves and relief valves representative of those in use or planned for use in PWR plants. To obtain valve performance data for the entire population of PWR plant valves, nine safety valves and ten relief valves were selected as a fully representative set of test valves. Justification that the selected valves represent all PWR plant valves was provided by each safety and relief valve manufacturer. Both the valve selection and justification work was performed as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of the PWR utilities in response to the recommendations of NUREG 0578 and the requirements of the NRC. Results of the Safety and Relief Valve Selection and Justification effort is documented in this report.

  7. Analysis of a bending test on a full-scale PWR hot leg elbow containing a surface crack

    Energy Technology Data Exchange (ETDEWEB)

    Delliou, P. le [Electricite de France, EDF, 77 - Moret-sur-Loing (France). Dept. MTC; Julisch, P.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Bezdikian, G. [Electricite de France, EDF, 92 - Paris la Defense (France). Direction Production Transport

    1998-11-01

    EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWR. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had undergone a 2400 hours ageing heat treatment at 400 C. The test preparation and execution, as well as the material characterization programme, were committed to MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200 C. For safety reasons, it took place on an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN.m), the notch did not initiate. This paper presents the experimental results and the fracture mechanics analysis of the test, based on finite element calculations. (orig.)

  8. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  9. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  10. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

  11. Development of laser weld monitoring system for PWR space grid

    International Nuclear Information System (INIS)

    The laser welding monitoring system was developed to inspect PWR space grid welding for KNFC. The demands for this optical monitoring system were applied to Q.C. and process control in space grid welding. The thermal radiation signal from weld pool can be get the variation of weld pool size. The weld pool size and depth are verified by analyzed wavelength signals from weld pool. Applied this monitoring system in space grid weld, improved the weld productivity. (author). 4 refs., 5 tabs., 31 figs

  12. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author)

  13. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    1997-02-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  14. Sizewell B - analysis of British application of US PWR technology

    International Nuclear Information System (INIS)

    This report provides information on the staff's evaluation of major design differences and issues developed by the British in their application (Sizewell B) of US PWR technology. One design change, the addition of steam-driven charging pumps, was assessed to have a relatively high value compared to the other changes. However, the assessment is based on a number of assumptions for which inadequate data exist to make an unqualified judgment. Other changes to the US design (as typified by the SNUPPS design) were found to have relatively low or moderate safety benefits for US application

  15. Conversion ratio in epithermal PWR, in thorium and uranium cycle

    International Nuclear Information System (INIS)

    Results obtained for the conversion ratio in PWR reactors with close lattices, operating in thorium and uranium cycles, are presented. The study of those reactors is done in an unitary fuel cell of the lattices with several ratios V sub(M)/V sub(F), considering only the equilibrium cycles and adopting a non-spatial depletion calculation model, aiming to simulate mass flux of reactor heavy elements in the reactor. The neutronic analysis and the cross sections generation are done with Hammer computer code, with one critical apreciation about the application of this code in epithermal systems and with modifications introduced in the library of basic data. (E.G.)

  16. Axial simulation of PWR core and study of actuators

    International Nuclear Information System (INIS)

    Development of an operation code allowing to simulate the behaviour of a PWR type reactor core. Load following is controled by bore and control rods, taking into account the temperature counter-reactions. The fine behaviour of the fuel element during transients is not simulated, on the other hand the central part of the reactor is completely simulated. The regulation equation are easily modifiable and thus it is possible to test in open loop any modification brought about to this regulation. Description of simulation tests on CAS-2B reactor: core control, static tests, dynamic tests

  17. Modelling of pellet-cladding interaction in PWR's

    International Nuclear Information System (INIS)

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  18. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  19. Experimental investigation and CFD validation of countercurrent flow limitation (CCFL) in a large-diameter hot-leg PWR geometry

    International Nuclear Information System (INIS)

    Counter current flow limitation CCFL is one of the phenomena that incorporate complex two-phase flows, including the existence of numerous flow patterns simultaneously, a complicated gas/liquid interface, and interfacial momentum transfer. Such a complexity makes it one of the challenging two-phase flow configurations for CFD validation. Numerous experimental investigations were carried out in recent years to enlarge the existing knowledge about this phenomenon. However, most of those investigations were carried out either in small-diameter geometry, or in a non-realistic geometry (rectangular cross section instead of a circular pipe). A review of experimental investigations shows that the scale and geometry have a large impact upon CCFL. In order to provide a better understanding of this phenomenon in a real PWR hot-leg geometry, and at a relatively large-diameter and scale, a test facility was constructed for this purpose. The facility consists of a reactor vessel simulator, a hot-leg geometry pipe (with 190 mm inner diameter), and a steam generator simulator. The facility represents a ∼1/3.9 scale of a PWR geometry and is completely made of transparent material allowing detailed optical observations. Experimental investigations were carried out at atmospheric pressure using distilled water and air. High-speed recording was implemented to acquire high-quality images of the air/water interface for experimental analysis and CFD validations. CCFL mechanisms, flow patterns, and the limits of the onset of CCFL and deflooding were experimentally identified. Current measurements are compared against previous investigations showing diverse effects of scale and geometry upon results. CFD simulations of two representative experimental cases were carried out and validated against the experimentally acquired air/water interface and the pressure difference between the reactor vessel and the steam generator. The CFD simulations shows the required improvements of this

  20. Caries inhibition by fluoride-releasing primers.

    Science.gov (United States)

    Kerber, L J; Donly, K J

    1993-10-01

    This study evaluated the caries inhibition of dentin primers with the addition of fluoride. Two standardized Class V preparations were placed in 20 molars, the gingival margin placed below the cementoenamel junction and the occlusal margin placed in enamel. Two dentin primers (Syntac and ScotchPrep) were placed in equal numbers of 20 preparations, according to manufacturer's instructions. Ammonium fluoride (10% by weight) was then added to these primers and they were placed in the remaining 20 preparations, opposing the non-fluoridated primer of the same system. All teeth were then restored with a non-fluoridated resin composite. All teeth were subjected to an artificial caries challenge (pH 4.2) for 5 days. Sections of 100 microns were obtained, photographed under polarized light microscopy, then demineralized areas were quantitated by digitization. Results demonstrated the mean areas (mm2 +/- S.D.) demineralization at 0.25 mm, 0.5 mm and 1.0 mm from the restoration margin to be: Syntac/fluoride (1.44 +/- 0.49, 1.68 +/- 0.54, 3.72 +/- 0.74); Syntac (1.99 +/- 0.58, 1.50 +/- 0.35, 2.98 +/- 1.26); ScotchPrep/fluoride (1.23 +/- 0.68, 1.55 +/- 0.64, 3.08 +/- 1.16); ScotchPrep (1.90 +/- 0.83, 1.71 +/- .038, 3.36 +/- 0.62). A paired t-test indicated primers with fluoride to demonstrate significantly less demineralization 0.25 mm from the restoration margin (P < 0.07). PMID:7880460

  1. Primer design using Primer Express® for SYBR Green-based quantitative PCR.

    Science.gov (United States)

    Singh, Amarjeet; Pandey, Girdhar K

    2015-01-01

    To quantitate the gene expression, real-time RT-PCR or quantitative PCR (qPCR) is one of the most sensitive, reliable, and commonly used methods in molecular biology. The reliability and success of a real-time PCR assay depend on the optimal experiment design. Primers are the most important constituents of real-time PCR experiments such as in SYBR Green-based detection assays. Designing of an appropriate and specific primer pair is extremely crucial for correct estimation of transcript abundance of any gene in a given sample. Here, we are presenting a quick, easy, and reliable method for designing target-specific primers using Primer Express(®) software for real-time PCR (qPCR) experiments.

  2. Loop-Mediated Amplification Accelerated by Stem Primers

    Directory of Open Access Journals (Sweden)

    Laurence Tisi

    2011-12-01

    Full Text Available Isothermal nucleic acid amplifications (iNAATs have become an important alternative to PCR for in vitro molecular diagnostics in all fields. Amongst iNAATs Loop-mediated amplification (LAMP has gained much attention over the last decade because of the simplicity of hardware requirements. LAMP demonstrates performance equivalent to that of PCR, but its application has been limited by the challenging primer design. The design of six primers in LAMP requires a selection of eight priming sites with significant restrictions imposed on their respective positioning and orientation. In order to relieve primer design constraints we propose an alternative approach which uses Stem primers instead of Loop primers and demonstrate the application of STEM-LAMP in assaying for Clostridium difficile, Listeria monocytogenes and HIV. Stem primers used in LAMP in combination with loop-generating and displacement primers gave significant benefits in speed and sensitivity, similar to those offered by Loop primers, while offering additional options of forward and reverse orientations, multiplexing, use in conjunction with Loop primers or even omission of one or two displacement primers, where necessary. Stem primers represent a valuable alternative to Loop primers and an additional tool for IVD assay development by offering more choices for primer design at the same time increasing assay speed, sensitivity, and reproducibility.

  3. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  4. SCOR 1000: an economic and innovative conceptual design PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M.; Chenaud, M.S. [CEA Cadarache (DEN/DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Tourniaire, B. [CEA Grenoble (DEN/DTN/SE2T/LPTM), 38 (France)

    2007-07-01

    Within the framework of innovative reactors studies, the Cea proposes the SCOR design (Simple COmpact Reactor) based on most of the advantages of innovative reactors. All main components are integrated in the vessel: the pressurizer, the canned pumps, the control rod mechanics of the driving system (CMD), and the dedicated heat exchangers of the passive heat removal system. The only steam generator is located above the vessel instead of the upper head. This design is featured by its compactness and by a large suppression or simplification of auxiliary systems. The first design with a 600 MWe shows its competitiveness with regard to the large loop-type PWR. To reduce the cost investment by the law sized effect, we examine the possibility of increasing the power of the reactor, while keeping the safety advantages of the medium sized SCOR. The electrical power of the new design is 1000 MWe. SCOR-1000 operates at much lower primary circuit pressure than standard PWRs (93 bars instead of the usual 155 bars), and the power density is lower (80 MW/m3 instead of 100 for the present PWRs). The reactivity is controlled by the CMD and by the burnable poison, without soluble boron. With the same safety advantages of the medium-sized SCOR, the cost reduction of the investment and of cost production could reach 18% with regard to the loop-type PWR. (authors)

  5. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  6. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)

  7. Degradation of fastener in reactor internal of PWR

    International Nuclear Information System (INIS)

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  8. VERA Core Simulator Methodology for PWR Cycle Depletion

    Energy Technology Data Exchange (ETDEWEB)

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  9. PWR reactor vessel in-service-inspection according to RSEM

    Energy Technology Data Exchange (ETDEWEB)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul [Intercontrole, 13, rue du Capricorne - SILIC 433, 94583 Rungis - Cedex (France)

    2006-07-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high

  10. Monte Carlo primer for health physicists

    International Nuclear Information System (INIS)

    The basic ideas and principles of Monte Carlo calculations are presented in the form of a primer for health physicists. A simple integral with a known answer is evaluated by two different Monte Carlo approaches. Random number, which underlie Monte Carlo work, are discussed, and a sample table of random numbers generated by a hand calculator is presented. Monte Carlo calculations of dose and linear energy transfer (LET) from 100-keV neutrons incident on a tissue slab are discussed. The random-number table is used in a hand calculation of the initial sequence of events for a 100-keV neutron entering the slab. Some pitfalls in Monte Carlo work are described. While this primer addresses mainly the bare bones of Monte Carlo, a final section briefly describes some of the more sophisticated techniques used in practice to reduce variance and computing time

  11. Project description: ORNL PWR blowdown heat transfer separate-effects program, Thermal-Hydraulic Test Facility (THTF)

    Energy Technology Data Exchange (ETDEWEB)

    1976-02-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results will be obtained from the Thermal-Hydraulic Test Facility (THTF), a large nonnuclear pressurized-water loop that incorporates a 49-rod electrically heated bundle. Supporting experiments will be carried out in two additional test loops - the Forced Convection Test Facility (FCTF), a small high-pressure facility in which single heater rods can be tested in annular geometry; and an air-water loop which is used to evaluate two-phase flow-measuring instrumentation.

  12. Effect of cyclic loadings on the stress corrosion crack growth rate in Alloy 600 in PWR primary water

    International Nuclear Information System (INIS)

    Fatigue air pre-cracked Compact Tensile (CT) specimens in Alloy 600 were tested in primary water (325 deg. C) of Pressurized Water Reactors (PWR). In order to assess the effect of cyclic loading on crack growth, CT specimens are tested under constant loadings and low frequencies cyclic loadings: triangular and saw-tooth. Two Alloy 600 materials, with different intrinsic susceptibility to Stress Corrosion Cracking (SCC), are studied. Crack growth rates are monitored in-situ by the direct current potential drop method and are validated by postmortem observations. Fracture surfaces are characterized by macroscopic and microscopic observations. Comparison of the crack growth rate and of the fracture features demonstrated that they depend on the characteristics of the mechanical loading (constant, triangular or sawtooth) and on the material intrinsic sensibility to SCC. (authors)

  13. Weathered Hydrocarbon Wastes: A Risk Management Primer

    OpenAIRE

    Brassington, Kirsty J.; Hough, Rupert L.; Paton, Graeme I.; Semple, Kirk T.; Risdon, Graeme C.; Crossley, Jane; Hay, I; Askari, K.; Pollard, Simon J. T.

    2007-01-01

    We provide a primer and critical review of the characterization, risk assessment, and bioremediation of weathered hydrocarbons. Historically the remediation of soil contaminated with petroleum hydrocarbons has been expressed in terms of reductions in total petroleum hydrocarbon (TPH) load rather than reductions in risk. There are several techniques by which petroleum hydrocarbons in soils can be characterized. Method development is often driven by the objectives of published...

  14. Primer for the algebraic geometry of sandpiles

    OpenAIRE

    Perkinson, David; Perlman, Jacob; Wilmes, John

    2011-01-01

    The Abelian Sandpile Model (ASM) is a game played on a graph realizing the dynamics implicit in the discrete Laplacian matrix of the graph. The purpose of this primer is to apply the theory of lattice ideals from algebraic geometry to the Laplacian matrix, drawing out connections with the ASM. An extended summary of the ASM and of the required algebraic geometry is provided. New results include a characterization of graphs whose Laplacian lattice ideals are complete intersection ideals; a new...

  15. Book review: feminist research practice: a primer

    OpenAIRE

    Smith, Emma

    2013-01-01

    The fully revised and updated Second Edition of Feminist Research Practice: A Primer, edited by Sharlene Nagy Hesse-Biber, draws on the expertise of a wide group of interdisciplinary scholars who aim to cover cutting-edge research methods and explore research questions related to the complex and diverse issues that deeply impact women’s lives. Emma Smith finds that it will be valuable for academics already working from or looking to develop their understanding and use of feminist research pra...

  16. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    International Nuclear Information System (INIS)

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

  17. Universal COI primers for DNA barcoding amphibians.

    Science.gov (United States)

    Che, Jing; Chen, Hong-Man; Yang, Jun-Xiao; Jin, Jie-Qiong; Jiang, Ke; Yuan, Zhi-Yong; Murphy, Robert W; Zhang, Ya-Ping

    2012-03-01

    DNA barcoding is a proven tool for the rapid and unambiguous identification of species, which is essential for many activities including the vouchering tissue samples in the genome 10K initiative, genealogical reconstructions, forensics and biodiversity surveys, among many other applications. A large-scale effort is underway to barcode all amphibian species using the universally sequenced DNA region, a partial fragment of mitochondrial cytochrome oxidase subunit I COI. This fragment is desirable because it appears to be superior to 16S for barcoding, at least for some groups of salamanders. The barcoding of amphibians is essential in part because many species are now endangered. Unfortunately, existing primers for COI often fail to achieve this goal. Herein, we report two new pairs of primers (➀, ➁) that in combination serve to universally amplify and sequence all three orders of Chinese amphibians as represented by 36 genera. This taxonomic diversity, which includes caecilians, salamanders and frogs, suggests that the new primer pairs will universally amplify COI for the vast majority species of amphibians.

  18. Avulsión de canino y subluxación del primer premolar superiores derechos

    OpenAIRE

    Berástegui, Esther

    1996-01-01

    Se presenta un caso clínico de avulsión de canino y subluxación del primer remolar, superiores derechos, secundario a traumatismo por piedra lanzada al aire. El tiempo extraoral fue de cuatro horas en medio seco. Se trató con reimplante, ferulización y tratamiento endodóncico. Los controles posteriores a los cuatro años del tratamiento mostraron la evolución favorable del caso.

  19. Benchmark exercise on SBLOCA experiment of PWR PACTEL facility

    International Nuclear Information System (INIS)

    Highlights: • PWR PACTEL, the facility with EPR type steam generators, is introduced. • The focus of the benchmark was on the analyses of the SBLOCA test with PWR PACTEL. • System codes with several modeling approaches were utilized to analyze the test. • Proper consideration of heat and pressure losses improves simulation remarkably. - Abstract: The PWR PACTEL benchmark exercise was organized in Lappeenranta, Finland by Lappeenranta University of Technology. The benchmark consisted of two phases, i.e. a blind and an open calculation task. Seven organizations from the Czech Republic, Germany, Italy, Sweden and Finland participated in the benchmark exercise, and four system codes were utilized in the benchmark simulation tasks. Two workshops were organized for launching and concluding the benchmark, the latter of which involved presentations of the calculation results as well as discussions on the related modeling issues. The chosen experiment for the benchmark was a small break loss of coolant accident experiment which was performed to study the natural circulation behavior over a continuous range of primary side coolant inventories. For the blind calculation task, the detailed facility descriptions, the measured pressure and heat losses as well as the results of a short characterizing transient were provided. For the open calculation task part, the experiment results were released. According to the simulation results, the benchmark experiment was quite challenging to model. Several improvements were found and utilized especially for the open calculation case. The issues concerned model construction, heat and pressure losses impact, interpreting measured and calculated data, non-condensable gas effect, testing several condensation and CCFL correlations, sensitivity studies, as well as break modeling. There is a clear need for user guidelines or for a collection of best practices in modeling for every code. The benchmark offered a unique opportunity to test

  20. CRISPR Primer Designer: Design primers for knockout and chromosome imaging CRISPR-Cas system.

    Science.gov (United States)

    Yan, Meng; Zhou, Shi-Rong; Xue, Hong-Wei

    2015-07-01

    The clustered regularly interspaced short palindromic repeats (CRISPR)-associated system enables biologists to edit genomes precisely and provides a powerful tool for perturbing endogenous gene regulation, modulation of epigenetic markers, and genome architecture. However, there are concerns about the specificity of the system, especially the usages of knocking out a gene. Previous designing tools either were mostly built-in websites or ran as command-line programs, and none of them ran locally and acquired a user-friendly interface. In addition, with the development of CRISPR-derived systems, such as chromosome imaging, there were still no tools helping users to generate specific end-user spacers. We herein present CRISPR Primer Designer for researchers to design primers for CRISPR applications. The program has a user-friendly interface, can analyze the BLAST results by using multiple parameters, score for each candidate spacer, and generate the primers when using a certain plasmid. In addition, CRISPR Primer Designer runs locally and can be used to search spacer clusters, and exports primers for the CRISPR-Cas system-based chromosome imaging system.

  1. CRISPR Primer Designer:Design primers for knockout and chromosome imaging CRISPR-Cas system

    Institute of Scientific and Technical Information of China (English)

    Meng Yan; Shi-Rong Zhou; Hong-Wei Xue

    2015-01-01

    The clustered regularly interspaced short palin-dromic repeats (CRISPR)-associated system enables biologists to edit genomes precisely and provides a powerful tool for perturbing endogenous gene regulation, modulation of epigenetic markers, and genome architecture. However, there are concerns about the specificity of the system, especial y the usages of knocking out a gene. Previous designing tools either were mostly built-in websites or ran as command-line programs, and none of them ran local y and acquired a user-friendly interface. In addition, with the development of CRISPR-derived systems, such as chromosome imaging, there were stil no tools helping users to generate specific end-user spacers. We herein present CRISPR Primer Designer for researchers to design primers for CRISPR applications. The program has a user-friendly interface, can analyze the BLAST results by using multiple parameters, score for each candidate spacer, and generate the primers when using a certain plasmid. In addition, CRISPR Primer Designer runs local y and can be used to search spacer clusters, and exports primers for the CRISPR-Cas system-based chromosome imaging system.

  2. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  3. Knowledge-based diagnosis of PWR secondary water chemistry

    International Nuclear Information System (INIS)

    A prototype knowledge-based diagnostic system has been developed for more effective processing of the in-line chemistry sensor data from the PWR secondary water-steam circuit with the SUN 3/80 workstation and the Nexpert Object shell program. The system consists of the data interface, the data interpreter, the CHEMISTRY-expert, the ACTION-expert, and the user interface. The knowledge base defines physical and conceptual models of the target domain in a class/object hierarchy, giving rise to a reduced number of rules with pattern matching. The rule base is broken down into separate rule groups for task control, classification, prioritization, and diagnosis to minimize the inference time. The system is scheduled for the Verification and Validation test to collect operational information feedback in one of the Korea nuclear power plants in the near future. (author)

  4. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P.J. [EPRI, Palo Alto, CA (United States)

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  5. PWR steam generator chemical cleaning, Phase I. Final report

    International Nuclear Information System (INIS)

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI

  6. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

  7. Design of an FPGA-based PWR ATWS mitigation system

    International Nuclear Information System (INIS)

    The present research is to explore the feasibility and conceptual design by using triple-redundant FPGA-based system for Anticipated-Transient-Without-Scram (ATWS) Mitigation System and Actuation Circuit (AMSAC) of a pressurized water reactor (PWR) type nuclear power plant (NPP). The Taipower's (Taiwan Power Company) Maanshan NPP was chosen for demonstration. An engineering simulated interface between AMSAC system and reactor/plant systems of Maanshan NPP was developed to provide an environment to validate the triple-redundant FPGA-based system. The software-free FPGA-based nuclear instrumentation and control (I and C) systems can easily be used for the modernization of the Taipower's nuclear power plant analog systems, thus may reduce the safety risk of undetectable software faults and common cause failures, and also minimize the regulatory licensing efforts and cost. (author)

  8. Development of high temperature adsorbent in PWR primary system

    International Nuclear Information System (INIS)

    Radiation exposure reduction in PWR is one of the most important problems to be solved. We have developed a high temperature Co adsorbent (HTA), which could be directly applied under primary reactor coolant conditions. This adsorbent was Fe-Ti-O system ceramics, and was fabricated to a suitable form for using in a packed column. Through those experiments of adsorption tests, compatibility tests, leaching tests and hot loop tests, it was found that HTA had superior adsorption capability to not only Co and Ni-ion but also many other transition metal ions. And it was also found that HTA was compatible with high temperature water, as well as advantageous for its waste solidification. Based on the experimental results, dose reduction effect was evaluated by a computer code. From this evaluation, it was found that more than 50 % dose reduction could be expected, when an advanced reactor coolant clean-up (RCC) system with HTA would be realized. (author)

  9. Sizewell 'B' PWR pre-construction safety report

    International Nuclear Information System (INIS)

    The Pre-Construction Safety Report (PCSR) for a PWR power station to be constructed as Sizewell 'B' is presented in 13 volumes containing 16 chapters. The PCSR has been submitted to the Nuclear Installations Inspectorate in support of the Central Electricity Generating Board's application for consent to the extension at Sizewell. It describes the design and provides the safety case for the proposed station, which comprises a 4-loop pressurized water reactor with associated generating plant and supporting auxiliary equipment. A general description of the station and its site is given. The strategy for ensuring nuclear safety is set out and the general design aspects of systems and plant outlined. The plant and systems, including their safety design bases and the fault analyses carried out for the design are described. Finally the way in which the plant will be decommissioned at the end of its useful life is outlined. (U.K.)

  10. Specification of water quality for the FRAMATOME PWR secondary circuit

    International Nuclear Information System (INIS)

    This paper describes the purpose, theory and scope of secondary system chemical specifications for FRAMATOME PWR nuclear power plants. All volatile treatment was chosen: controlling the feedwater pH by means of a volatile amine (ammonia, morpholine), and excluding oxygen by the addition of hydrazine. The pollutants are monitored at the steam generator drains by completely automatic measurements using simple and reliable techniques: pH measurement and a diagram of the cation conductivity versus sodium. An explanation is given of the monitoring techniques and to the effect of the various kinds of possible pollutant. A new concept is described, the annual quota expressed in day.microsiements.cm-1 which enables the amount of absorbed pollutants in the steam generator to be evaluated. The methods used for maintaining the desired chemical quality are dealt with

  11. Non linear identification applied to PWR steam generators

    International Nuclear Information System (INIS)

    For the precise industrial purpose of PWR nuclear power plant steam generator water level control, a natural method is developed where classical techniques seem not to be efficient enough. From this essentially non-linear practical problem, an input-output identification of dynamic systems is proposed. Through Homodynamic Systems, characterized by a regularity property which can be found in most industrial processes with balance set, state form realizations are built, which resolve the exact joining of local dynamic behaviors, in both discrete and continuous time cases, avoiding any load parameter. Specifically non-linear modelling analytical means, which have no influence on local joined behaviors, are also pointed out. Non-linear autoregressive realizations allow us to perform indirect adaptive control under constraint of an admissible given dynamic family

  12. PWR steam generator chemical cleaning, Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Rothstein, S.

    1978-07-01

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  13. Design of large steam turbines for PWR power stations

    International Nuclear Information System (INIS)

    The authors review the thermodynamic cycle requirements for use with pressurized-water reactors, outline the way thermal efficiency is maximized, and discuss the special nature of the wet-steam cycle associated with turbines for this type of reactor. Machine and cycle parameters are optimized to achieve high thermal efficiency, particular attention being given to arrangements for water separation and steam reheating and to provisions for feedwater heating. Principles and details of mechanical design are considered for a range both of full-speed turbines running at 3000 rev/min on 50 Hz systems and of half-speed turbines running at 1800 rev/min on 60 Hz systems. The importance of service experience with nuclear wet-stream turbines, and its relevance to the design of modern turbines for PWR applications, is discussed. (author)

  14. PWR pressurizer discharge piping system on-site testing

    International Nuclear Information System (INIS)

    Framatome PWR systems includes the installation of safety valves and relief valves wich permit the discharge of steam from the pressurizer to the pressurizer relief tank through discharge piping system. Water seal expulsion pluration then depends on valve stem lift dynamics which can vary according to water-stem interaction. In order to approaches the different phenomenons, it was decided to perform a test on a 900 MWe French plant, test wich objectives are: characterize the mechanical response of the discharge piping to validate a mechanical model; open one, two or several valves among the following: one safety valve and three pilot operated relief valves, at a time or sequentially and measure the discharge piping transient response, the support loads, the

  15. CFD application to PWR subchannel void distribution benchmark

    International Nuclear Information System (INIS)

    A CFD study is performed to simulate the steady-state void distribution benchmark based on the NUPEC PWR Subchannel and Bundle Tests (PSBT). The CFD calculation predicted the void distributions in central typical and thimble subchannels, side subchannel and corner subchannel. The CFD prediction shows a higher void fraction near the heated wall and a migration of void in the subchannel gap region. A measured image of void distribution indicated a locally higher void fraction near the heated wall. The CFD predictions of void fraction and fluid density agree well with the measured ones for the low void test condition. However, the CFD calculations tend to underpredict the void fraction and overpredict the fluid density as the void fraction increases. (author)

  16. A system for trip analysis of PWR reactors using neural networks

    International Nuclear Information System (INIS)

    This work presents the basic concepts and the general description of a computational system developed for trip analysis in PWR nuclear power plants which is based on neural networks and artificial intelligence concepts. (author)

  17. N4 PWR makes full use of distributed processing and local networks

    Energy Technology Data Exchange (ETDEWEB)

    Aschenbrenner, J.F.; Tetreau, F.; Colling, J.M.

    1988-01-01

    The new instrumentation and control systems for the French N4 PWR power plant make extensive use of programmable controllers based on advanced microprocessor technology and distributed processing. Local networking techniques are widely used which simplify architecture and equipment design.

  18. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    International Nuclear Information System (INIS)

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  19. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  20. Maintenance of Ni-based alloy at PWR plant

    International Nuclear Information System (INIS)

    Kansai Electric owns 11 PWR plants. At our PWR plants, we are taking various preventive maintenance measures on Ni-based alloy according to the prediction of possible trouble while past trouble occurred at overseas plants due to Primary Water Stress Corrosion Cracking (PWSCC) being considered. In addition, we are making an effort to put new maintenance techniques into practical use by conducting demonstration tests to confirm their applicability to actual plants. We have replaced reactor vessel heads at 7 plants with new ones. At the other 4 plants, we took, measures to reduce the temperature of reactor vessel head top to delay the timing of PWSCC occurrence. We are carrying out the constant load tests to predict the timing of PWSCC occurrence at these 4 plants. It is planned to conduct non-destructive inspections at an appropriate timing based on the result of the prediction. Based on the prediction of the timing of PWSCC occurrence at bottom-mounted instrumentation (BMI), we have developed water jet peening (WJP) technique to reduce residual stress and applied the technique to our plants successively. Meanwhile, a technique to cut and eliminate cracking has been developed. In addition, capping technique, which covers overall the concerned nozzle on the outer surface of the reactor vessel, has been also established. For alloy 132/82 weld metal for the connection, we are conducting ultrasonic inspection at our plants successively. In order to prepare against PWSCC occurrence, we have also established a technique to replace the entire section of concerned short piping with new one. (author)

  1. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  2. Westinghouse Passive Plants - AP600 and S PWR

    International Nuclear Information System (INIS)

    The original thought behind the AP600 passive design was that if the U. S. nuclear industry was to be revitalized, it would require a new, advanced technology with clearly proven benefits in safety. Response from the international arena indicates that, regardless of local domestic consideration, a revitalization of the U. S. industry is seen as very important, even essential, worldwide. And the potential for scale up of these passive safety features has been clearly established, allowing the benefits of the passive technology to be realized in countries that, for whatever reason, are interested in larger plant sizes only. Government projections indicate that U. S. energy demands in the 1990s will grow steadily, creating the need for approximately 117,000 to 322,000 MW of new generating capacity by the year 2010. Although this growth in electricity demand continues to be strong, orders for new nuclear power plants have not kept pace, in part due to licensing delays, prohibitive construction costs, and public uncertainty about safety. However, with the increased concerns about the environmental and economic security risks involved with an excessive dependence on fossil fuels, there is a growing realization that nuclear power must play a major role in our energy future. Looking to the future, Westinghouse is developing the AP600, a simplified two-loop PWR featuring passive safety systems. Drawing on the results of the AP600 development and testing programs, Westinghouse is also developing the larger S PWR, a passive, three-loop power plant with an output in the 900 to 1000 MW range

  3. First application of hollow fiber filter for PWR condensate polishing

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, S. [ORGANO Corp., Tokyo (Japan); Otoha, K.; Takiguchi, H. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    In Tsuruga Unit-2 (PWR 1160 MWe commenced commercial operation in 1987), current procedure for secondary system clean-up before start-up had prolonged outage time and had consumed a huge amount of de-ionized (DI) water. In addition, iron oxide in condensate had accelerated the degradation of condensate demineralizer (CD) resin. The corrosion product of iron could also influence the secondary side corrosion of steam generator (SG) tubing if it intruded into SG through CD. To solve these problems, Japan Atomic Power Company (JAPC) decided to introduce hollow fiber filter (HFF) type condensate filter into Tsuruga-2, as the first application to PWR in the world. Because of retro-fitted HFF in Tsuruga Unit-2, limitations for installation space and flow resistance in condensate system and cost reduction required new design for compact and low differential pressure system and for long life filter module. JAPC and ORGANO assessed methodologies to achieve these goals. An advanced HFF system, including a newly developed compact HFF module design, was installed at Tsuruga Unit-2 in 1997 based on the assessment. During the 5 years since the installation, the HFF system has provided excellent crud removal that enables to shorten the outage period and to reduce DI water consumption drastically. Stable differential pressure (dP) trend of the HFF system indicates an expected module life of more than 7 years, with backwash cleaning required only 2 or 3 times per year. In addition to providing the expected operating cost reduction and improved SG tube integrity, numerous additional benefits have resulted from the retrofit. (authors)

  4. Identifying thermal cycling mechanisms in PWR branch line piping

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Keller, J.D.; Bilanin, A.J. [Continuum Dynamics, Inc., Ewing, NJ (United States)

    2002-07-01

    Predicting the onset and the characteristics of thermal cycling in pressurized water reactor (PWR) branch line piping systems is critical to formulation of thermal fatigue screening tools. The complex nature of the underlying thermal-hydraulic phenomena, however, significantly complicates prediction using analytical models or direct numerical simulations. Instead, it is necessary to perform scaled experiments to identify the physical mechanisms and to gather data for formulation of semi-empirical models for the thermal cycling phenomena. Through the EPRI Materials Reliability Program a test program is underway to identify and develop semi-empirical correlations for the physical thermalhydraulic mechanisms that cause thermal cycling in dead-ended PWR branch line piping systems. Three series of tests are being performed in this test program: configuration tests on a representative up-horizontal (UH) branch line piping geometry, configuration tests on a representative down-horizontal (DH) branch line piping geometry, and high Reynolds number tests to assess penetration of secondary flow structures into a dead-ended branch line. Results from UH and DH configuration tests indicate that random turbulence penetration is not sufficient for thermal cycling to occur. Rather a swirling flow structure, representative of a large, 'corkscrew' vortical structure, is required for thermal cycling. Scale tests on the UH configuration have simulated cycling phenomena observed in full-scale plant data and have been used to determine parametric sensitivities in formulating a predictive model for the thermal cycling. Data indicate that the mechanism for thermal cycling in UH configurations is stochastic but scales with the leak rate from the valve. The critical dependent variables are reduced to several non-dimensional scaling curves, resulting in a semiempirical predictive model. This paper discusses the test program and the results obtained to date. Application of these

  5. An Experience on RCS CRUD Sampling in European PWR Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Jae; Kim, Jong Bin [Sung Woo E and T Co., Seoul (Korea, Republic of); Kang, Duk Won [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    In most PWRs the normal method of corrosion product sampling is to collect a 'grab' sample from either the RCS hot or cold leg. This method is not ideal and the results are often dominated by soluble and particulate transients that can bias them high by factors of between ten and one hundred times. Nevertheless 'grab' samples can still give relatively satisfactory results from which qualitative trends of total soluble plus particulate corrosion product concentrations can be determined and, although 'grab' sampling may not be ideal, it is useful in detecting and following abnormal particulate releases from the core. It is possible to eliminate the worst of the transient effects by collecting a sample from continuously flowing RCS sample line, but the changes necessary to operate in this way are major, will be costly and may not be practicable for many existing plants. The evaluation of changes in corrosion product concentrations, particularly when the changes increase the particulate concentrations, can indicate that there is a risk that an Axial Offset Anomaly (AOA) may develop, or a risk of increased corrosion product releases when the plant shuts down for refueling. Recently, Diablo Canyon and Callaway in United States America, Ringhals in Sweden, Sizewell B in Great Britain, Vandellos in Spain and Doel in Belgium, these PWR plants have applied capillary sampling method to CRUD Analysis in parallel with grab sampling method under the recommendation of EPRI. In this thesis, it will show the practice based on actually tested method in European PWR plants.

  6. Estimating PWR fuel rod failures throughout a cycle

    International Nuclear Information System (INIS)

    A fuel performance engineer requires good prediction models for fuel conditions to help assure that any fuel repair operation he may recommend for the next refueling outage will have a minimal impact on nuclear plant operation. For nearly two decades, simple equilibrium equations have been used to provide estimates of the number of failed fuel rods in a pressurized water reactor (PWR) core. The unknown parameter is the isotopic escape rate (upsilon), which is often assumed to be --1 X 10/sup -8//s for the release of /sup 131/I from a 3- to 4-m-long PWR rod. The use of this escape rate value will generally produce end-of-cycle (EOC) predictions that are accurate within a factor of --3. When applied at the time when fuel rods initially fail, such as early in a reactor cycle, however, the prediction obtained may overestimate the number of failed rods present by a factor of 10 or more. While a goal of Combustion Engineering's (C-E's) efforts on failed fuel prediction (FFP) models over the past decade has been to increase the accuracy of the EOC estimate, recent efforts have emphasized improving prediction capability for failed rods present early in a reactor cycle. The C-E approach to modeling iodine release from failed fuel rods is based on dynamic escape rate theory that is incorporated in the C-E IODYNE (for iodine dynamic evaluation) code. This theory has been empirically modified to account for specific observed time dependencies of the release rates for /sup 131/I and /sup 133/I from a failed rod. In a current version of IODYNE, four such factors have been included in the FFP model, as described in this paper

  7. MPprimer: a program for reliable multiplex PCR primer design

    Directory of Open Access Journals (Sweden)

    Wang Xiaolei

    2010-03-01

    Full Text Available Abstract Background Multiplex PCR, defined as the simultaneous amplification of multiple regions of a DNA template or multiple DNA templates using more than one primer set (comprising a forward primer and a reverse primer in one tube, has been widely used in diagnostic applications of clinical and environmental microbiology studies. However, primer design for multiplex PCR is still a challenging problem and several factors need to be considered. These problems include mis-priming due to nonspecific binding to non-target DNA templates, primer dimerization, and the inability to separate and purify DNA amplicons with similar electrophoretic mobility. Results A program named MPprimer was developed to help users for reliable multiplex PCR primer design. It employs the widely used primer design program Primer3 and the primer specificity evaluation program MFEprimer to design and evaluate the candidate primers based on genomic or transcript DNA database, followed by careful examination to avoid primer dimerization. The graph-expanding algorithm derived from the greedy algorithm was used to determine the optimal primer set combinations (PSCs for multiplex PCR assay. In addition, MPprimer provides a virtual electrophotogram to help users choose the best PSC. The experimental validation from 2× to 5× plex PCR demonstrates the reliability of MPprimer. As another example, MPprimer is able to design the multiplex PCR primers for DMD (dystrophin gene which caused Duchenne Muscular Dystrophy, which has 79 exons, for 20×, 20×, 20×, 14×, and 5× plex PCR reactions in five tubes to detect underlying exon deletions. Conclusions MPprimer is a valuable tool for designing specific, non-dimerizing primer set combinations with constrained amplicons size for multiplex PCR assays.

  8. Development of the Combination Method of PWR Spent Fuel for DUPIC Fuel Preparation

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Ju Ho; Kim, S. K.; Jung, T. C.; June, T. H.; Lee, J. M.; Kim, I. S.; Park, C. S.; Kim, M. J.; An, J. I.; Park, S. H. [Kyung Hee University, Seoul (Korea, Republic of)

    1997-07-15

    Optimum finding method of PWR spent fuel was developed in application of DUPIC fuel composition to nuclear fuel production. In order to make the database of the PWR spent fuel for the optimum composition, composition data of the PWR spent fuels from Youngkwang unit 1 and 2, Kori unit 3 and 4 and Uljin unit 1 and 2 were collected, analyzed and stored. Artificial intelligent access was attempted in optimizing the composition, and the combination algorithm for PWR spent fuel was developed. In this work database of the composition data of the PWR spent fuels from Youngkwang unit 1 and 2, Kori unit 3 and 4 and Uljin unit 1 and 2 as well as their combination algorithm for PWR spent fuel were developed. The combination algorithm is to find the combination of the spent fuel assembly which is quite close to the requirement per unit mass of DUPIC fuel. The required data are total weight of the fuel, tolerance of the errors, importance of the elements and the discharge data. This combination algorithm enables to find the optimum PWR spent fuel assembly for DUPIC fuel with the database of the spent fuels according to the DUPIC fuel standards. The combination algorithm developed in this work can afford the technical support to fuel supply in preparing the DUPIC fuel, and make contribution in DUPIC fuel cycle technology. It can be directly used in DUPIC fuel cycle technology, and can be also used in the management of the spent fuels with respect to their compositions and ingredients as well as the nuclear safeguards. Composition of the PWR spent fuel in each assembly depends on the initial concentration, degree of combustion, specific power, and its location in the reactor core. They may be affected by the kind of fuel rod and its axial length. Therefore, analysis procedures in these regards should be established for the effective application of the results of this work. 6 refs., 12 tabs., 46 figs. (author)

  9. Bias identification in PWR pressurizer instrumentation using the generalized liklihood-ratio technique

    International Nuclear Information System (INIS)

    A method for detecting and identifying biases in the pressure and level sensors of a pressurized water reactor (PWR) pressurizer is described. The generalized likelihood ratio (GLR) technique performs statistical tests on the innovations sequence of a Kalman filter state estimator and is capable of determining when a bias appears, in what sensor the bias exists, and estimating the bias magnitude. Simulation results using a second-order linear, discrete PWR pressurizer model demonstrate the capabilities of the GLR method

  10. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  11. Identification and evaluation of PWR in-vessel severe accident management strategies

    International Nuclear Information System (INIS)

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

  12. Bayesian models a statistical primer for ecologists

    CERN Document Server

    Hobbs, N Thompson

    2015-01-01

    Bayesian modeling has become an indispensable tool for ecological research because it is uniquely suited to deal with complexity in a statistically coherent way. This textbook provides a comprehensive and accessible introduction to the latest Bayesian methods-in language ecologists can understand. Unlike other books on the subject, this one emphasizes the principles behind the computations, giving ecologists a big-picture understanding of how to implement this powerful statistical approach. Bayesian Models is an essential primer for non-statisticians. It begins with a definition of probabili

  13. Signals and systems primer with Matlab

    CERN Document Server

    Poularikas, Alexander D

    2006-01-01

    Signals and Systems Primer with MATLAB® equally emphasizes the fundamentals of both analog and digital signals and systems. To ensure insight into the basic concepts and methods, the text presents a variety of examples that illustrate a wide range of applications, from microelectromechanical to worldwide communication systems. It also provides MATLAB functions and procedures for practice and verification of these concepts.Taking a pedagogical approach, the author builds a solid foundation in signal processing as well as analog and digital systems. The book first introduces orthogonal signals,

  14. Primer on CDM programme of activities

    Energy Technology Data Exchange (ETDEWEB)

    Hinostroza, M. (UNEP Risoe Centre, Roskilde (Denmark)); Lescano, A.D. (A2G Carbon Partners (Peru)); Alvarez, J.M. (Ministerio del Ambiente del Peru (Peru)); Avendano, F.M. (EEA Fund Management Ltd. (United Kingdom)

    2009-07-01

    As an advanced modality introduced in 2005, the Programmatic CDM (POA) is expected to address asymmetries of participation, especially of very small-scale project activities in certain areas, key sectors and many countries with considerable potential for greenhouse gas emission reductions, not reached by the traditional single-project-based CDM. Latest experiences with POAs and the recently finalized official guidance governing the Programmatic CDM are the grassroots of this Primer, which has the purpose of supporting the fully understanding of rules and procedures of POAs by interpreting them and analyzing real POA cases. Professional and experts from the public and private entities have contributed to the development of this Primer, produced by the UNEP Risoe Centre, as part of knowledge support activities for the Capacity Development for the CDM (CD4CDM) project. The overall objective of the CD4CDM is to develop the capacities of host countries to identify, design, approve, finance, implement CDM projects and commercialize CERs in participating countries. The CDM4CDM is funded by the Netherlands Ministry of Foreign Affairs. (author)

  15. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  16. pmoA Primers for Detection of Anaerobic Methanotrophs▿

    OpenAIRE

    Luesken, F.A.; Zhu, B.; Alen, T.A. van; Butler, M.K.; Diaz, M. R.; Song, B.; Op den Camp, H.J.M.; M. S. M. Jetten; Ettwig, K.F.

    2011-01-01

    Published pmoA primers do not match the pmoA sequence of “Candidatus Methylomirabilis oxyfera,” a bacterium that performs nitrite-dependent anaerobic methane oxidation. Therefore, new pmoA primers for the detection of “Ca. Methylomirabilis oxyfera”-like methanotrophs were developed and successfully tested on freshwater samples from different habitats. These primers expand existing molecular tools for the study of methanotrophs in the environment.

  17. The chemical decontamination of the Callisto PWR loop

    International Nuclear Information System (INIS)

    The CALLISTO (Capability for Light water Irradiation in Steady state and Transient Operation) is a PWR experimental facility for scientific in-pile studies installed into the BR2 Material Test Reactor. Three experimental rigs, called In-Pile Sections (IPS), are installed in three reactor channels. They are connected to a common pressurized loop, which operates with representative PWR water chemistry (typically 400 ppm boron, 3,5 ppm lithium and 30 ccSTP/kg dissolved hydrogen). The IPSs can be provided with adequate instrumentation and be modified to perform valid irradiation studies in a high neutron flux and in a relevant thermos-hydraulic environment. During more than 15 years of operation, activation products have accumulated into the loop leading to a continuous increase of the dose rates at the work area. Consequently periodic maintenance and inspection operations have become more and more expensive in terms of collective dose uptake. In consultation with the internal and external safety authorities the decision has been made to proceed to the chemical closed-loop decontamination of the most important components of CALLISTO (heater, pressurizer, main and bleed flow coolers). The objective of reducing the dose rates without compromising the integrity of the operational loop has led to the combined use of known soft chemical decontamination products as KMnO4 and H2C2O4. About 10 GBq of Co-60 activity and 250 g of corrosion products were removed from the stainless steel CALLISTO loop. The systems involved had a total volume of 0,5 m3 and a surface area of 18 m2. All released activity and corrosion products were removed by ion exchange resins, leading to the generation of 2x150 liters of radioactive waste. The dose rate reduction factors in contact with the treated components varied between 2 and 12. The collective dose uptake of the entire operation (preparation - decontamination - clean-up) was about 5,5 man.mSv, and thereby in line with the ALARA estimations

  18. Issues and remedies for secondary system of PWR/VVER

    International Nuclear Information System (INIS)

    Secondary side degradation of steam generators (SG) and Flow Accelerated Corrosion (FAC) in the secondary system have been for a long time important issues in PWR and VVER types of Nuclear Power Plants. With the evolution of the design, the most important issues are progressively moving from secondary side corrosion of Alloy 600 SG tubing, which is being replaced, to a larger variety of risks associated with potential inadequate chemistries. As far as FAC of carbon steel is concerned, the evolution of treatment selection for minimizing corrosion products transport toward the SG, as well as progressive replacement of components in the feedwater train, decreases the risk of dramatic failures which have occurred in the past. After having briefly explained the reason for the past problems encountered in the secondary system of PWR and VVER, this paper evaluates the risk associated with various impurities or contaminants that may be present in the secondary system and how to mitigate them in the most appropriate, efficient, economical and environmental friendly way. The covered species are sodium, calcium, magnesium, chloride, sulfate and sulfur compounds, fluorides, organic compounds, silica, oxygen, lead, ion exchange resins. This paper also proposes the best remedies for mitigating the new issues that may be encountered in operating plants or units under construction. These are mainly: - Selecting a steam water treatment able to minimize the quantity of corrosion products transported toward the SG; - Mitigating the risk of Flow Induced Vibration by a proper control of deposits in sensitive areas; - Minimizing the risk of concentration of impurities in local areas where they may induce corrosion; - Avoiding the presence of abnormal quantities of some species in SG, such as the detrimental presence of lead and ion exchange resin debris or the controversial presence of organic compounds; - Optimizing costs of maintenance activities (SG mechanical, chemical cleaning

  19. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  20. High-speed measurement of firearm primer blast waves

    OpenAIRE

    Courtney, Michael; Daviscourt, Joshua; Eng, Jonathan; Courtney, Amy

    2012-01-01

    This article describes a method and results for direct high-speed measurements of firearm primer blast waves employing a high-speed pressure transducer located at the muzzle to record the blast pressure wave produced by primer ignition. Key findings are: 1) Most of the lead styphnate based primer models tested show 5.2-11.3% standard deviation in the magnitudes of their peak pressure. 2) In contrast, lead-free diazodinitrophenol (DDNP) based primers had standard deviations of the peak blast p...

  1. High universality of matK primers for barcoding gymnosperms

    Institute of Scientific and Technical Information of China (English)

    Yan LI; Lian-Ming GAO; RAM C.POUDEL; De-Zhu Li; Alan FORREST

    2011-01-01

    DNA barcoding is a tool to provide rapid and accurate taxonomic identification using a standard DNA region. A two-marker combination of rnatK+rbcL was formally proposed as the core barcode for land plants by the Consortium for the Barcode of Life Plant Working Group. However, there are currently no barcoding primers for matK showing high universality in gymnosperms. We used 57 gymnosperm species representing 40 genera, 11families and four subclasses to evaluate the universality of nine candidate matK primers and one rbcL primer in this study. Primer (1F/724R) of rbcL is proposed here as a universal primer for gymnosperms due to high universality. One of the nine candidate matK primers (Gym_F1A/Gym_R1A) is proposed as the best "universal" matK primer for gynnosperms because of high polymerase chain reaction success and routine generation of high quality bidirectional sequences. A specific matK primer for Ephedra was newly designed in this study, which performed well on the sampled species. The primers proposed here for rbcL and matK can be easily and successfully amplified for most gymnosperms.

  2. GENOMEMASKER package for designing unique genomic PCR primers

    Directory of Open Access Journals (Sweden)

    Kaplinski Lauris

    2006-03-01

    Full Text Available Abstract Background The design of oligonucleotides and PCR primers for studying large genomes is complicated by the redundancy of sequences. The eukaryotic genomes are particularly difficult to study due to abundant repeats. The speed of most existing primer evaluation programs is not sufficient for large-scale experiments. Results In order to improve the efficiency and success rate of automatic primer/oligo design, we created a novel method which allows rapid masking of repeats in large sequence files, for example in eukaryotic genomes. It also allows the detection of all alternative binding sites of PCR primers and the prediction of PCR products. The new method was implemented in a collection of efficient programs, the GENOMEMASKER package. The performance of the programs was compared to other similar programs. We also modified the PRIMER3 program, to be able to design primers from lowercase-masked sequences. Conclusion The GENOMEMASKER package is able to mask the entire human genome for non-unique primers within 6 hours and find locations of all binding sites for 10 000 designed primer pairs within 10 minutes. Additionally, it predicts all alternative PCR products from large genomes for given primer pairs.

  3. Evaluation of conservation programs: a primer

    Energy Technology Data Exchange (ETDEWEB)

    Soderstrom, E.J.; Berry, L.G.; Hirst, E.; Bronfman, B.H.

    1981-07-01

    This primer is an introduction to the field of program evaluation. A considerable amount of information is included about the actual conduct of an evaluation. No specific activity is addressed, but rather an idea is provided of the issues and constraints characteristic of each phase and activity of an evaluation. Examples of how these issues arise and potential solutions are presented. Chapter 1 provides background on the origins and evolution of evaluation, focusing especially on why evaluation is currently receiving increased emphasis. Chapter 2 is an overview of the field of evaluation, especially as it relates to energy conservation. Chapter 3 is a consideration of the steps involved in conducting an evaluation from the planning stage to data analysis. Chapter 4 discusses managerial considerations in conducting an evaluation. Chapter 5 discusses evaluators as people; both in terms of organizational obstacles likely to be encountered by, and personal characteristics found in, good evaluators. An appendix includes suggested further readings.

  4. Complex and Adaptive Dynamical Systems A Primer

    CERN Document Server

    Gros, Claudius

    2011-01-01

    We are living in an ever more complex world, an epoch where human actions can accordingly acquire far-reaching potentialities. Complex and adaptive dynamical systems are ubiquitous in the world surrounding us and require us to adapt to new realities and the way of dealing with them. This primer has been developed with the aim of conveying a wide range of "commons-sense" knowledge in the field of quantitative complex system science at an introductory level, providing an entry point to this both fascinating and vitally important subject. The approach is modular and phenomenology driven. Examples of emerging phenomena of generic importance treated in this book are: -- The small world phenomenon in social and scale-free networks. -- Phase transitions and self-organized criticality in adaptive systems. -- Life at the edge of chaos and coevolutionary avalanches resulting from the unfolding of all living. -- The concept of living dynamical systems and emotional diffusive control within cognitive system theory. Techn...

  5. Complex and adaptive dynamical systems a primer

    CERN Document Server

    Gros, Claudius

    2013-01-01

    Complex system theory is rapidly developing and gaining importance, providing tools and concepts central to our modern understanding of emergent phenomena. This primer offers an introduction to this area together with detailed coverage of the mathematics involved. All calculations are presented step by step and are straightforward to follow. This new third edition comes with new material, figures and exercises. Network theory, dynamical systems and information theory, the core of modern complex system sciences, are developed in the first three chapters, covering basic concepts and phenomena like small-world networks, bifurcation theory and information entropy. Further chapters use a modular approach to address the most important concepts in complex system sciences, with the emergence and self-organization playing a central role. Prominent examples are self-organized criticality in adaptive systems, life at the edge of chaos, hypercycles and coevolutionary avalanches, synchronization phenomena, absorbing phase...

  6. Complex and adaptive dynamical systems a primer

    CERN Document Server

    Gros, Claudius

    2015-01-01

    This primer offers readers an introduction to the central concepts that form our modern understanding of complex and emergent behavior, together with detailed coverage of accompanying mathematical methods. All calculations are presented step by step and are easy to follow. This new fourth edition has been fully reorganized and includes new chapters, figures and exercises. The core aspects of modern complex system sciences are presented in the first chapters, covering network theory, dynamical systems, bifurcation and catastrophe theory, chaos and adaptive processes, together with the principle of self-organization in reaction-diffusion systems and social animals. Modern information theoretical principles are treated in further chapters, together with the concept of self-organized criticality, gene regulation networks, hypercycles and coevolutionary avalanches, synchronization phenomena, absorbing phase transitions and the cognitive system approach to the brain. Technical course prerequisites are the standard ...

  7. Primer for criticality calculations with DANTSYS

    International Nuclear Information System (INIS)

    With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his or her facility. Typically, two types of codes are available: deterministic codes such as ANISN or DANTSYS that solve an approximate model exactly and Monte Carlo Codes such as KENO or MCNP that solve an exact model approximately. Often, the analyst feels that the deterministic codes are too simple and will not provide the necessary information, so most modeling uses Monte Carlo methods. This sometimes means that hours of effort are expended to produce results available in minutes from deterministic codes. A substantial amount of reliable information on nuclear systems can be obtained using deterministic methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico in cooperation with the Radiation Transport Group at Los Alamos National Laboratory has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. (DANTSYS is the name of a suite of codes that users more commonly know as ONEDANT, TWODANT, TWOHEX, and THREEDANT.) It assumes a college education in a technical field, but there is no assumption of familiarity with neutronics codes in general or with DANTSYS in particular. The primer is designed to teach by example, with each example illustrating two or three DANTSYS features useful in criticality analyses

  8. Formaldehyde as hypothetical primer of biohomochirality

    Energy Technology Data Exchange (ETDEWEB)

    Goldanskii, V.I. [N. N. Semenov Institute of Chemical Physics of the Russian Academy of Sciences, Kosygin Street 4, Moscow, 117334 (Russia)

    1996-07-01

    One of the most intriguing and crucial problems of the prebiotic evolution and the origin of life is the explanation of the origin of biohomochirality. A scheme of conversions originated by formaldehyde (FA) as hypothetical primer of biohomochirality is proposed. The merit of FA as executor of this function is based -inter alia - on the distinguished role of FA as one of the earliest and simplest molecules in both warm, terrestrial and cold, extraterrestrial scenarios of the origin of life. The confirmation of the role of FA as primer of biohomochirality would support the option of an RNA world as an alternative to the protein world. The suggested hypothesis puts forward for the first time a concrete sequence of chemical reactions which can lead to biohomochirality. The spontaneous breaking of the mirror symmetry is secured by the application of the well-known Frank scheme (combination of autocatalysis and {open_quote}{open_quote}annihilation{close_quote}{close_quote} of L and D enantiomers) to the series of interactions of FA {open_quote}{open_quote}trimers{close_quote}{close_quote} (i.e. C{sub 3}H{sub 6}O{sub 3} compounds) of (aaa), (apa) and (app) types, where the monomeric groups (a) means {open_quote}{open_quote}achirons{close_quote}{close_quote} (a=CH{sub n}, n{ge}2 and C=M, M=C,O) and (p) mean {open_quote}{open_quote}prochirons{close_quote}{close_quote} (p=HC{asterisk}OM, M=H,C). {copyright} {ital 1996 American Institute of Physics.}

  9. Reassessment of PWR pressure-vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    A continuing analysis of the PTS problem associated with PWR postuated OCA's indicates that the previously accepted degree of conservatism in the fracture-mechanics model needs to be more closely evaluated, and if excessive, reducted. One feature that was believed to be conservative was the use of two-dimensional as opposed to finite-length (three-dimensional) flaws. A flaw of particular interest is one that is located in an axial weld of a plate-type vessel. For those vessels that suffer relatively high radiation damage in the welds, the length of the flaw will be no greater than the length of the weld, and recent calculations indicate that a deep flaw of that length (approx. 2 m) is not effectively infinitely long, contrary to previous thinking. The benefit to be derived from consideration of the 2-m flaw and also a semielliptical flaw with a length-to-depth ratio of 6/1 was investigated by analyzing several postulated transients. In doing so the sensitivity of the benefit to a specified maximum crack arrest toughness and to the duration of the transient was investigated. Results of the analysis indicate that for some conditions the benefit in using the 2-m flaw is substantial, but it decreases with increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the limit on crack arrest toughness

  10. Applicability of oxygenated water chemistry for PWR secondary systems

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden); Takiguchi, H.; Otoha, K. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    Introduction of oxygenated water chemistry (OWC) in PWR secondary side is considered as a means to reduce the transportation of corrosion products into the steam generator and thus also minimizing crevice deposits and subsequent materials problems. One main concern, however, is the risk of inter-granular attack (IGA) in crevices. In order to study effects on crevice tube IGA by OWC, a series of experiments were performed in a steam generator (SG) simulating loop. This comprised a SG tube and a tube support plate (TSP) together forming the crevice. The over-all objective of the work accounted here was to demonstrate that it is possible to operate the steam generator secondary side with OWC without causing intolerable IGA or other types of attack on the tube in the crevice area. Tubes of sensitized Alloy 600 were exposed during a total of nine experiments in an autoclave using a TSP/tube arrangement with an asymmetric crevice design. Experiments were performed at high and low pH and potential under open and packed crevice conditions. The aggressiveness of the crevice environment was also further increased by addition of carbonate and chloride. Furthermore the tube was pressurized. Experimental parameters were monitored on the primary side as well as in the secondary bulk phase and in the crevice. (authors)

  11. Enhancing heat transfer and crud mitigation in PWR fuel

    International Nuclear Information System (INIS)

    This paper discusses three methods for increasing single phase heat transfer in PWR fuel. The primary effect of increasing heat transfer is a reduction in the steaming rate from the fuel rods, which in turn reduces the likelihood of crud formation on the fuel rods and the potential for adsorption of boron into the crud. The advantage of lowering boron mass on the fuel is reduced risk of Axial Offset Anomaly (AOA). Another benefit of reduced crud formation is a lower risk of localized corrosion, a known contributor to rod cladding failures. Thinner crud leads to locally lower rod operating temperatures (lower corrosion rate) since crud acts as a thermal insulator between the rod and the coolant. The first method of increasing heat transfer involves addition of more than one Intermediate Flow Mixing vane grid (IFM) in the span between two neighboring structural spacing grids. The second method includes optimization of the mixing vane according to axial position. The third method involves variation of the IFMs axial position to optimize axial distribution of rod heat transfer. (authors)

  12. Boron mixing transient in a PWR vessel. Physical studies

    International Nuclear Information System (INIS)

    EDF has conducted a R and D action, aiming at gaining more knowledge on vessel thermal-hydraulics; it consists of two complementary approaches based on mock-up experiments and numerical simulations. Maintenance scenarios studies began in 1995. They have been performed solely with the FEM CFD code N3S. The FEM model take into account the U pipe, the primary pump and the cold leg. This mesh can be connected to the vessel mesh used in the study of previous configurations. The first case in progress concerns the influence of the start-up of a boron unsaturated demineralizer. The study concerns the plug formation in the U pipe involved by the clear and cold seal injection water entering the primary circuit. At the end of the diluted water injection the primary pump is started up and the U pipe fluid is sent in the reactor vessel. This paper presents first the CPY 900 MW PWR vessel taken into account in these physical studies, with a special focus on the geometric peculiarities. Then the 1/5. scale BORA-BORA mock-up and the 3D FEM Thermal Hydraulic code N3S are described. The results obtained until now are presented. The degree of achievement of the studies on the three priority cases (start-up, hot shut-down normal operation, cold shut-down normal operation)

  13. PWR loading pattern optimization using Harmony Search algorithm

    International Nuclear Information System (INIS)

    Highlights: ► Numerical results reveal that the HS method is reliable. ► The great advantage of HS is significant gain in computational cost. ► On the average, the final band width of search fitness values is narrow. ► Our experiments show that the search approaches the optimal value fast. - Abstract: In this paper a core reloading technique using Harmony Search, HS, is presented in the context of finding an optimal configuration of fuel assemblies, FA, in pressurized water reactors. To implement and evaluate the proposed technique a Harmony Search along Nodal Expansion Code for 2-D geometry, HSNEC2D, is developed to obtain nearly optimal arrangement of fuel assemblies in PWR cores. This code consists of two sections including Harmony Search algorithm and Nodal Expansion modules using fourth degree flux expansion which solves two dimensional-multi group diffusion equations with one node per fuel assembly. Two optimization test problems are investigated to demonstrate the HS algorithm capability in converging to near optimal loading pattern in the fuel management field and other subjects. Results, convergence rate and reliability of the method are quite promising and show the HS algorithm performs very well and is comparable to other competitive algorithms such as Genetic Algorithm and Particle Swarm Intelligence. Furthermore, implementation of nodal expansion technique along HS causes considerable reduction of computational time to process and analysis optimization in the core fuel management problems

  14. Robots in P.W.R. nuclear powerplants

    International Nuclear Information System (INIS)

    The satisfactory operation of 37 900-MWe PWR powerplants in France, Belgium and South-Africa and the start-up of 1300 MWe powerplants allowed the development of a wide range of automatic units and robots for the periodic maintenance of nuclear plants, reducing the risk of ionizing radiation for the personnel. A large number of automated tools have been built. Among them: - inspection and maintenance systems for the tube bundle of steam generators, - robotized arms ROTETA and ROMEO for the heavy maintenance and delicate operations such as tube extraction or shot peening of tubes to improve their resistance to corrosion; - the versatile manipulator T.A.M. with electrically controlled articulations. The development of functionally versatile tools and robots and the integration of new technologies such as 3-D vision allowed the construction of the self-guided vehicle FRASTAR capable of moving within a nuclear building and in a cluttered environment. This vehicle includes means for avoiding isolated obstacles and can move on stairs

  15. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  16. Integrated training support system for PWR operator training simulator

    International Nuclear Information System (INIS)

    The importance of operator training using operator training simulator has been recognized intensively. Since 1986, we have been developing and providing many PWR simulators in Japan. We also have developed some training support systems connected with the simulator and the integrated training support system to improve training effect and to reduce instructor's workload. This paper describes the concept and the effect of the integrated training support system and of the following sub-systems. We have PES (Performance Enhancement System) that evaluates training performance automatically by analyzing many plant parameters and operation data. It can reduce the deviation of training performance evaluation between instructors. PEL (Parameter and Event data Logging system), that is the subset of PES, has some data-logging functions. And we also have TPES (Team Performance Enhancement System) that is used aiming to improve trainees' ability for communication between operators. Trainee can have conversation with virtual trainees that TPES plays automatically. After that, TPES automatically display some advice to be improved. RVD (Reactor coolant system Visual Display) displays the distributed hydraulic-thermal condition of the reactor coolant system in real-time graphically. It can make trainees understand the inside plant condition in more detail. These sub-systems have been used in a training center and have contributed the improvement of operator training and have gained in popularity. (author)

  17. Dynamic modelling of PWR fuel assembly for seismic behaviour

    International Nuclear Information System (INIS)

    Vibration and snap back tests have shown that the behaviour of PWR fuel assemblies was non linear : the fuel assembly eigenfrequencies decrease with the excitation level or with the motion amplitude, which was supposed to be due to the slippage of the fuel rods through the grids. Up to now the fuel assembly models were linear and composed by one beam alone representing both the guide thimbles and the fuel rods or by two beams (one for the guide thimbles and one for the fuel rods). The stiffness of such models' were adjusted to fit with the measured eigenfrequency corresponding to a given amplitude. The aim of this paper is to identify the influence of the slippage between grids and fuel rods on the dynamic behaviour of the fuel assembly. For that purpose a non linear fuel assembly model is proposed representing explicitly the slippage phenomenon and is applied to the reduced scale fuel assemblies which have been tested in the framework of a collaboration between FRAMATOME and CEA-DMT. Comparisons between calculations and experiments will be presented and the limitation of this model will be also discussed

  18. Break location effects on PWR small break LOCA phenomena

    International Nuclear Information System (INIS)

    The report presents experimental results of a small lower plenum break test of SB-PV-01 conducted at the large-Scale Test Facility (LSTF) of the Rig-of-Safety Assessment (ROSA)-IV program. This test simulates a loss-of-coolant accident (LOCA) caused by instrument tubes break (break area corresponds to 0.5% of the cold leg flow area) in a Westinghouse-type pressurized water reactor (PWR) assuming both manual actuation for all of the high pressure injection (HPI) systems and failure of the auxiliary feedwater systems. The report clarifies long-term system responses, especially the core cooling conditions related to the primary mass inventory. Also it clarifies break location effects on small break LOCA phenomena by comparing other five similar LOCA tests with break locations at cold leg, hot leg, upper head, pressurizer top (TMI-type) and SG U-tubes. It is coucluded that the lower plenum break is the severest on core heatup due to the highest break flow rate and the least primary mass recovery after the ECCS among the six tests. (author)

  19. Automatic defect identification on PWR nuclear power station fuel pellets

    International Nuclear Information System (INIS)

    This article presents a new automatic identification technique of structural failures in nuclear green fuel pellet. This technique was developed to identify failures occurred during the fabrication process. It is based on a smart image analysis technique for automatic identification of the failures on uranium oxide pellets used as fuel in PWR nuclear power stations. In order to achieve this goal, an artificial neural network (ANN) has been trained and validated from image histograms of pellets containing examples not only from normal pellets (flawless), but from defective pellets as well (with the main flaws normally found during the manufacturing process). Based on this technique, a new automatic identification system of flaws on nuclear fuel element pellets, composed by the association of image pre-processing and intelligent, will be developed and implemented on the Brazilian nuclear fuel production industry. Based on the theoretical performance of the technology proposed and presented in this article, it is believed that this new system, NuFAS (Nuclear Fuel Pellets Failures Automatic Identification Neural System) will be able to identify structural failures in nuclear fuel pellets with virtually zero error margins. After implemented, the NuFAS will add value to control quality process of the national production of the nuclear fuel.

  20. Manufacture of nuclear fuel elements for commercial PWR in China

    International Nuclear Information System (INIS)

    Yibin Nuclear Fuel Element Plant (YFP) under the leadership of China National Nuclear Corporation is sole manufacturer in China to specialize in the production of fuel assemblies and associated core components for commercial PWR nuclear power plant. At the early of 1980's, it began to manufacture fuel assemblies and associated core components for the first core of QINSHAN 300 MW nuclear power plant designed and built by China itself. With the development of nuclear power industry in China and the demand for localization of nuclear fuel elements in the early 1990's, YFP cooperated with FRAMATOME France in technology transfer for design and manufacturing of AFA 2G fuel assembly and successfully supplied the qualified fuel assemblies for the reloads of two units of GUANGDONG Da Ya Bay 900 MW nuclear power plant (Da Ya Bay NPP), and has achieved the localization of fuel assemblies and nuclear power plants. Meanwhile, it supplied fuel assemblies and associated core components for the first core and further reloads of Pakistan CHASHMA 300 MW nuclear power plant which was designed and built by China, and now it is manufacturing AFA 2G fuel assemblies and associated core components for the first core of two units of NPQJVC 600 MW nuclear power plant. From 2001 on, YFP will be able to supply Da Ya Bay NPP with the third generation of fuel assembly-AFA 3G which is to realize a strategy to develop the fuel assembly being of long cycle reload and high burn-up

  1. Effect of coolant chemistry on PWR radiation transport processes

    International Nuclear Information System (INIS)

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. While the extent of in-core spinel deposition is influenced by pH in a manner to be expected from the temperature coefficient of solubility of nickel-iron spinel, there is evidence that boric acid plays a role apart from its influence on pH. Out-of-core deposition of active cobalt on stainless steel takes place largely in the chromium-rich inner oxide layer, and there is also significant uptake of corrosion products into the film on Zircaloy. Deposition depends on flow characteristics in different ways for different elements. The evidence suggests that in DWL soluble species are dominant in out-of-core deposition processes for corrosion products. The adsorption of cobalt in zirconium oxide provides a route for deposition on fuel elements which may in some circumstances be more significant than spinel deposition. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. Some alternative chemistry regimes have been explored, but do not appear to offer any advantages with respect to activity transport control over the more conventional regime based on lithium hydroxide and hydrogen dosing. 8 refs., 26 figs., 28 tabs

  2. PWR-440 water chemistry optimization to reduce AOA effect

    International Nuclear Information System (INIS)

    The pressure drop increase in PWR-440 is mainly caused by the fact that the coolant contains numerous corrosion products, which are generated after decontamination and deposited in the top part of the fuel assembly as well as by coolant nucleate boiling that under standard water chemistry conditions leads to acceleration of corrosion products deposition and coolant radioactivity growth respectively. The modeling of the pressure drop changes were based on standard data of water chemistry, reactor operating characteristics and fundamental thermodynamic parameters to predict the pressure drop growth. The results of the performed research and modeling of the corrosion products mass transfer processes allowed to qualify relative contribution of thermohydraulic and chemical parameters in the processes and to fulfill the activities as follows: To perform power units operation at water chemistry with maximum permissible alkali metals content. To increase the coolant flow rate through the core; to do so, throttling orifices were replaced and canister-shields were removed. To reduce the number of steam generators to be decontaminated to 2 per year in a single power unit. As a result deposits accumulation in fuel assemblies has been minimized and there is no leakage in the fuel element; reactor thermal output limitation has been eliminated. (author)

  3. Barium silicate glass/Inconel X-750 interaction. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kelsey, Jr., P. V.; Siegel, W. T.; Miley, D. V.

    1980-01-01

    Water reactor safety programs at the Idaho National Engineering Laboratory have required the development of specialized instrumentation. An example is the electrical conductivity-sensitive liquid level transducer developed for use in pressurized-water reactors (PWRs) in which the operation of the sensing probe relies upon the passage of current through the water between the center pin of the electrode and its shell such that when water is present the resulting voltage is low, and conversely, when water is absent the voltage is high. The transducer's ceramic seal is a hot-pressed glass ceramic; its metal housing is Inconel X-750. The ceramic material provides an essential dielectric barrier between the center pin and the outer housing. The operation of the probe as well as the integrity of the PWR environment requires a hermetically-bonded seal between the ceramic and the metal. However, during testing, an increasing number of probe assemblies failed owing to poor glass-to-metal seals as well as void formation within the ceramic. Therefore, a program was initiated to characterize the metallic surface with respect to pre-oxidation treatment and determine optimum conditions for wetting and bonding of the metal by the glass to obtain baseline data relevant to production of acceptable transducer seals.

  4. PWR safety/relief valve blowdown analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)

    1982-10-01

    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.

  5. Advanced methods for the study of PWR cores

    International Nuclear Information System (INIS)

    This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in October 2003. The transparencies of the annual meeting are presented in the introductive part: 1 - status of the French nuclear park: nuclear energy results, management of an exceptional climatic situation: the heat wave of summer 2003 and the power generation (J.C. Barral); 2 - status of the research on controlled thermonuclear fusion (J. Johner). Then follows the technical session about the advanced methods for the study of PWR reactor cores: 1 - the evolution approach of study methodologies (M. Lambert, J. Pelet); 2 - the point of view of the nuclear safety authority (D. Brenot); 3 - the improved decoupled methodology for the steam pipe rupture (S. Salvatores, J.Y. Pouliquen); 4 - the MIR method for the pellet-clad interaction (renovated IPG methodology) (E. Baud, C. Royere); 5 - the improved fuel management (IFM) studies for Koeberg (C. Cohen); 6 - principle of the methods of accident study implemented for the European pressurized reactor (EPR) (F. Foret, A. Ferrier); 7 - accident studies with the EPR, steam pipe rupture (N. Nicaise, S. Salvatores); 8 - the co-development platform, a new generation of software tools for the new methodologies (C. Chauliac). (J.S.)

  6. [Analysis of effectiveness of cDNA synthesis, induced using complementary primers and primers containing a noncomplementary base matrix].

    Science.gov (United States)

    D'iachenko, L B; Chenchik, A A; Khaspekov, G L; Tatarenko, A O; Bibilashvili, R Sh

    1994-01-01

    We have studied the efficiency of DNA synthesis catalyzed by M-MLV reverse transcriptase or Thermus aquaticus DNA polymerase for primers (4-17 nucleotides long) either completely matched or possessing a single mismatched base pair at all possible positions in the primer. It has been shown that DNA synthesis efficiency depends not only on the position of mismatched base pair but on the length and primary structure of the primer. The enzyme, template, and primer concentrations determine the relative level of mismatched DNA synthesis.

  7. Effect of ethanolamine injection on wall thinning rate of PWR carbon steel components

    International Nuclear Information System (INIS)

    For pH control of PWR secondary system water chemistry, some plants have changed to ethanolamine injection. The purpose of this work was to understand the effect of changing water chemistry on wall thinning rate of the PWR secondary system due to flow accelerated corrosion. For that purpose, evaluations of water chemistry were carried out by a mass balance calculation, wall thinning rate measurement by a rotary disk test and wall thinning rate evaluation based on calculated magnetite solubility. As a result, it was found to be effective to inhibit the wall thinning rate of the PWR secondary system due to flow accelerated corrosion by ethanolamine injection, but it was not sufficiently effect to neglect wall thinning rate due to flow accelerated corrosion. The effect of the wall thinning rate inhibition also varied greatly for each component of the PWR secondary system. It was found that maintenance of the carbon steel used for the PWR secondary system was still required under ethanolamine injection condition. (author)

  8. Effect of thione primers on adhesive bonding between an indirect composite material and Ag-Pd-Cu-Au alloy.

    Science.gov (United States)

    Imai, Hideyuki; Koizumi, Hiroyasu; Shimoe, Saiji; Hirata, Isao; Matsumura, Hideo; Nikawa, Hiroki

    2014-01-01

    The current study evaluated the effect of primers on the shear bond strength of an indirect composite material joined to a silverpalladium-copper-gold (Ag-Pd-Cu-Au) alloy (Castwell). Disk specimens were cast from the alloy and were air-abraded with alumina. Eight metal primers were applied to the alloy surface. A light-polymerized indirect composite material (Solidex) was bonded to the alloy. Shear bond strength was determined both before and after the application of thermocycling. Two groups primed with Metaltite (thione) and M. L. Primer (sulfide) showed the greatest post-thermocycling bond strength (8.8 and 6.5 MPa). The results of the X-ray photoelectron spectroscopic (XPS) analysis suggested that the thione monomer (MTU-6) in the Metaltite primer was strongly adsorbed onto the Ag-Pd-Cu-Au alloy surface even after repeated cleaning with acetone. The application of either the thione (MTU-6) or sulfide primer is effective for enhancing the bonding between a composite material and Ag-Pd-Cu-Au alloy.

  9. Survey of experiments and code development for the passive residual heat removal system of PWR in China

    Institute of Scientific and Technical Information of China (English)

    HUANG Yan-Ping; ZHUO Wen-Bing; YANG Zu-Mao; XIAO Ze-Jun; CHEN Bing-De; JIA Dou-Nan

    2004-01-01

    Three different kinds of experiments and their typical results are surveyed for the passive residual heat removal system (PRHRS) of PWR performed in Nuclear Power Institute of China (NPIC) recent ten years. The typical results of MISAP. a special code for PWR passive residual heat removal system developed and assessed by NPIC,are also described briefly in this paper.

  10. Polymerase chain reaction of Au nanoparticle-bound primers

    Institute of Scientific and Technical Information of China (English)

    SHEN Hebai; HU Min; YANG Zhongnan; WANG Chen; ZHU Longzhang

    2005-01-01

    Polymerase chain reaction (PCR) is a useful technique for in vitro amplification of a DNA fragment. In this paper, a PCR procedure using Au nanoparticle (AuNP) -bound primers was systemically studied. The 5′-SH- (CH2)6-modified primers were covalently attached to the AuNP surface via Au-S bonds, and plasmid pBluescript SK was used as a template. The effects of the concentration of AuNP-bound primers, annealing temperature and PCR cycles were evaluated, respectively. The results indicate that PCR can proceed successfully under optimized condition, with either forward or reverse primers bound to the AuNP surface or with both the two primers bound to the AuNP surface. Development of PCR procedure based on AuNPs not only makes the isolation of PCR products very convenient, but also provides novel methods to prepare AuNP-bound ssDNA and nanostructured material.

  11. Uncoupling primer and releaser responses to pheromone in honey bees

    Science.gov (United States)

    Grozinger, Christina M.; Fischer, Patrick; Hampton, Jacob E.

    2007-05-01

    Pheromones produce dramatic behavioral and physiological responses in a wide variety of species. Releaser pheromones elicit rapid responses within seconds or minutes, while primer pheromones produce long-term changes which may take days to manifest. Honeybee queen mandibular pheromone (QMP) elicits multiple distinct behavioral and physiological responses in worker bees, as both a releaser and primer, and thus produces responses on vastly different time scales. In this study, we demonstrate that releaser and primer responses to QMP can be uncoupled. First, treatment with the juvenile hormone analog methoprene leaves a releaser response (attraction to QMP) intact, but modulates QMP’s primer effects on sucrose responsiveness. Secondly, two components of QMP (9-ODA and 9-HDA) do not elicit a releaser response (attraction) but are as effective as QMP at modulating a primer response, downregulation of foraging-related brain gene expression. These results suggest that different responses to a single pheromone may be produced via distinct pathways.

  12. Summary Day 1: Second AirMonTech Workshop, Current and Future Air Quality Monitoring

    OpenAIRE

    Querol, Xavier

    2012-01-01

    Presentación resumen de las ponencias del primer día de las Second AirMonTech Workshop, Current and Future Air Quality Monitoring. Estas jornadas tuvieron lugar en Barcelona del 25 al 26 de abril de 2012.

  13. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Putney, J.M.; Preece, R.J. [National Power, Leatherhead (GB). Technology and Environment Centre

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3.

  14. Development of a lead extrusion damper for PWR reactor coolant loop system

    International Nuclear Information System (INIS)

    Conventional seismic design for PWR reactor coolant loop system is conducted under a philosophy of rigid design and large site of rigid supports and many snubbers are used as seismic supports. But recently various type of alternative supports to snubbers have been proposed. A lead extrusion damper (LED) is one of the devices being considered. This paper is devoted to experimental and analytical work on the development of the LED for PWR reactor coolant loop system. In the study, the fundamental mechanism of the damper and the damping effect on the response of a steam generator supported by the LED were studied. From experimental and analytical approaches, the feasibility of application of the LED to PWR reactor coolant loop system was confirmed

  15. CFD studies on the phenomena around counter-current flow limitations of gas/liquid two-phase flow in a model of a PWR hot leg

    Energy Technology Data Exchange (ETDEWEB)

    Deendarlianto, E-mail: deendarlianto@ugm.ac.id [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Department of Mechanical and Industrial Engineering, Faculty of Engineering, Gadjah Mada University, Jalan Grafika No. 2, Yogyakarta 55281 (Indonesia); Hoehne, Thomas; Lucas, Dirk; Vallee, Christophe [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Zabala, Gustavo Adolfo Montoya [Department of Chemical Engineering, Simon Bolivar University, Valle of Sartenejas, Caracas 1080 (Venezuela, Bolivarian Republic of)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer We modelled CCFL in a PWR hot leg using Algebraic Interfacial Area Density model. Black-Right-Pointing-Pointer The model is able to distinguish the local flow morphologies. Black-Right-Pointing-Pointer Test fluids are air-water and steam-water. Black-Right-Pointing-Pointer Calculated CCFL and water level are in good agreement with experimental data. - Abstract: In order to improve the understanding of counter-current two-phase flow and to validate new physical models, CFD simulations of a 1/3rd scale model of the hot leg of a German Konvoi pressurized water reactor (PWR) with rectangular cross section were performed. Selected counter-current flow limitation (CCFL) experiments conducted at Helmholtz-Zentrum Dresden-Rossendorf (HZDR) were calculated with ANSYS CFX using the multi-fluid Euler-Euler modelling approach. The transient calculations were carried out using a gas/liquid inhomogeneous multiphase flow model coupled with a shear stress transport (SST) turbulence model. In the simulation, the drag law was approached by a newly developed correlation of the drag coefficient in the Algebraic Interfacial Area Density (AIAD) model. The model can distinguish the bubbles, droplets and the free surface using the local liquid phase volume fraction value. A comparison with the high-speed video observations shows a good qualitative agreement. The results indicate also a quantitative agreement between calculations and experimental data for the CCFL characteristics and the water level inside the hot leg channel.

  16. UniFrag and GenomePrimer : selection of primers for genome-wide production of unique amplicons

    NARCIS (Netherlands)

    van Hijum, SAFT; de Jong, A; Buist, G; Kok, J; Kuipers, OP

    2003-01-01

    The complementary programs UniFrag and GenomePrimer were developed to provide a reliable high-throughput method to select the most unique regions within genomic DNA sequence(s) and design primers therein, involving minimal user intervention and maximum flexibility.

  17. PrimerSeq:Design and Visualization of RT-PCR Primers for Alternative Splicing Using RNA-seq Data

    Institute of Scientific and Technical Information of China (English)

    Collin Tokheim; Juw Won Park; Yi Xing

    2014-01-01

    The vast majority of multi-exon genes in higher eukaryotes are alternatively spliced and changes in alternative splicing (AS) can impact gene function or cause disease. High-throughput RNA sequencing (RNA-seq) has become a powerful technology for transcriptome-wide analysis of AS, but RT-PCR still remains the gold-standard approach for quantifying and validating exon splicing levels. We have developed PrimerSeq, a user-friendly software for systematic design and visualization of RT-PCR primers using RNA-seq data. PrimerSeq incorporates user-provided tran-scriptome profiles (i.e., RNA-seq data) in the design process, and is particularly useful for large-scale quantitative analysis of AS events discovered from RNA-seq experiments. PrimerSeq features a graphical user interface (GUI) that displays the RNA-seq data juxtaposed with the expected RT-PCR results. To enable primer design and visualization on user-provided RNA-seq data and transcript annotations, we have developed PrimerSeq as a stand-alone software that runs on local computers. PrimerSeq is freely available for Windows and Mac OS X along with source code at http://primerseq.sourceforge.net/. With the growing popularity of RNA-seq for transcriptome stud-ies, we expect PrimerSeq to help bridge the gap between high-throughput RNA-seq discovery of AS events and molecular analysis of candidate events by RT-PCR.

  18. Alternative water chemistry for the primary loop of PWR plants

    International Nuclear Information System (INIS)

    Advanced fuel element concepts (longer cycles, higher burnup, increased rod power) call for more reactivity binding capacity and, moreover, might produce higher void fractions, particularly in the hot channel. Thus, on the one hand, more alcalizing agent is needed to maintain a high coolant pH according to the approved ''modified boron-lithium mode of operation'' in the presence of more boric acid (chemical shim); on the other hand, increasing enrichment of coolant constituents due to local boiling (higher void fraction), which must not result in accelerated corrosion of fuel cladding and structural materials, imposes enhanced requirements on both, materials technology and water chemistry. At present, the use of boric acid enriched in B10 (the isotope effective in terms of reactivity control) appears to advantageously compromise in capturing more neutrons with less total boron while maintaining or even slightly reducing lithium concentrations at the same time. There is no feasible alternative for boric acid used as the chemical shim and for hydrogen gas as the reducing agent used to suppress oxygen formation by water radiolysis. Systematic screening as performed in phase 1 of a recent project proved potassium hydroxide to be the only potential candidate to favourably replace lithium 7 hydroxide as an alcalizing agent. Unfortunately, the results of pertinent comparative corrosion tests are not unambiguous, and available operational experience with potassium hydroxide in WWER plants is not readily applicable to western world-type PWR plants. Therefore, a switch-over from lithium to potassium can be envisaged only subsequent to a comprehensive qualification program which is planned to be the objective of phase 2 of the project. This program should also comprise zinc addition tests in order to confirm the alleged positive impact of this element on corrosion rates and activity buildup. Supplementary, it is recommended to consider amendments to existing water chemistry

  19. Analyses of PWR boron dilution consequences with the Arrotta code

    Energy Technology Data Exchange (ETDEWEB)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced.

  20. Effect of water chemistry on deposition for PWR plant operation

    International Nuclear Information System (INIS)

    For Pressurized Water Reactor (PWR) operation, water chemistry guidelines, specifications and associated surveillance programs are key to avoid deposition of oxides. Deposition of oxides can be detrimental by disrupting results of flow measurements, decreasing the thermal exchange capacity, or even by impairing safety. This paper describes the most important cases of deposition, their consequences for operation, and the implemented improvements to avoid their reoccurrence. Deposition that led to a Crud Induced Power Shift (CIPS) is also described. In the primary and in the secondary sides, orifice plates are typically used for measuring feedwater flow rate in nuclear power plants. Feedwater flow rates are used for control purposes and are important safety parameters as they are used to determine the plant's operating power level. Fouling of orifice plates in the primary side has been found during surveillance testing. For reactor coolant pumps, the formation of deposits on the seal No.1 can cause abnormally high or low leak rates through the seal. The leak rate through this seal must be carefully maintained within a prescribed range during plant operation. In the secondary side, orifice plate fouling has been the cause of feedwater flow/reference thermal power drift. For the steam generators (SG), magnetite deposition has led to fouling of the tube bundle, clogging of the quadri-foiled support plate holes and hard sludge formation on the base plate. For the generators, copper hollow conductors are widely used. Buildup of copper oxides on the interior walls of copper conductors has caused insufficient heat transfer. All these deposition cases have received adequate attention, understanding and response via improvement of our surveillance programs. (authors)

  1. Primer on electricity futures and other derivatives

    Energy Technology Data Exchange (ETDEWEB)

    Stoft, S.; Belden, T.; Goldman, C.; Pickle, S.

    1998-01-01

    Increased competition in bulk power and retail electricity markets is likely to lower electricity prices, but will also result in greater price volatility as the industry moves away from administratively determined, cost-based rates and encourages market-driven prices. Price volatility introduces new risks for generators, consumers, and marketers. Electricity futures and other derivatives can help each of these market participants manage, or hedge, price risks in a competitive electricity market. Futures contracts are legally binding and negotiable contracts that call for the future delivery of a commodity. In most cases, physical delivery does not take place, and the futures contract is closed by buying or selling a futures contract on or near the delivery date. Other electric rate derivatives include options, price swaps, basis swaps, and forward contracts. This report is intended as a primer for public utility commissioners and their staff on futures and other financial instruments used to manage price risks. The report also explores some of the difficult choices facing regulators as they attempt to develop policies in this area.

  2. Primer on electricity futures and other derivatives

    International Nuclear Information System (INIS)

    Increased competition in bulk power and retail electricity markets is likely to lower electricity prices, but will also result in greater price volatility as the industry moves away from administratively determined, cost-based rates and encourages market-driven prices. Price volatility introduces new risks for generators, consumers, and marketers. Electricity futures and other derivatives can help each of these market participants manage, or hedge, price risks in a competitive electricity market. Futures contracts are legally binding and negotiable contracts that call for the future delivery of a commodity. In most cases, physical delivery does not take place, and the futures contract is closed by buying or selling a futures contract on or near the delivery date. Other electric rate derivatives include options, price swaps, basis swaps, and forward contracts. This report is intended as a primer for public utility commissioners and their staff on futures and other financial instruments used to manage price risks. The report also explores some of the difficult choices facing regulators as they attempt to develop policies in this area

  3. Evaluation of PWR steam generator water hammer. Final technical report, June 1, 1976--December 31, 1976

    International Nuclear Information System (INIS)

    An investigation of waterhammer in the main feedwater piping of PWR steam generators due to water slugs formed in the steam generator feedring is reported. The relevant evidence from PWR operation and testing is compiled and summarized. The state-of-the-art of analysis of related phenomena is reviewed. Original exploratory modeling experiments at 1/10 and 1/4 scale are reported. Bounding analyses of the behavior are performed and several key phenomena have been identified for the first time. Recommendations to the Nuclear Regulatory Commission are made

  4. Application of diffusion theory methods to PWR [pressurized water reactors] analysis

    International Nuclear Information System (INIS)

    In-core physics analysis of pressurized light water reactors (PWRs) requires accurate predictions of three-dimensional pin-by-pin power distributions. The PWR analyses must rely on diffusion theory approximation because no practical methods exist for performing routine three-dimensional pin-by-pin transport calculations. Pin-by-pin diffusion calculations are also prohibitively expensive in three-dimensional geometry, and PWR analyses utilize either two-dimensional pin-by-pin models or three-dimensional advanced nodal models. The purpose of this paper is to detail and contrast approximations required by pin-by-pin and nodal diffusion methods

  5. Contribution to the study of the conversion PWR type reactors to the thorium cycle

    International Nuclear Information System (INIS)

    The use of the thorium cycle in PWR reactors is discussed. The fuel has been calculated in the equilibrium condition for a economic comparison with the uranium cycle (in the same condition). First of all, a code named EQUILIBRIO has been developed for the fuel equilibrium calculation. The results gotten by this code, were introduced in the LEOPARD code for the fuel depletion calculation (in the equilibrium cycle). Same important physics details of fuel depletion are studied, for instance: the neutron balance, power sharing, fuel burnup, etc. The calculations have been done taking as reference the Angra-1 PWR reactor. (Author)

  6. Assessment of options for the treatment of Sizewell PWR liquid effluent

    International Nuclear Information System (INIS)

    This report describes the origins of PWR liquid waste streams, their composition and rates of arising. Data has been collected from operational PWRs and estimates obtained for Sizewell B PWR liquid waste streams. Current liquid waste treatment practices are reviewed and assessments made of established and novel treatment techniques which could be applicable to Sizewell B. A short list of treatment options is given and recommendations are made relating to established treatment technologies suitable for Sizewell B and also to development work on more novel treatments which could lead to a reduction in waste disposal volumes. (author)

  7. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  8. Behavior of a PWR-containment under rising internal pressure load

    International Nuclear Information System (INIS)

    Reactor safety containments are dimensioned so that in a postulated design accident (fracture of the primary duct), the coolant flowing out can be reliably accommodated. The internal pressure in German PWR's is about 5 bar. Radioactive contamination of the environment is largely avoided in this way. The experiments were done for the safety containment of the PWR at Philippsburg. It is a freestanding spherical shell 56 metres in diameter and 38 mm thick. The tensioning of the concrete foundations is 400 below the equator. The spherical shell is welded from about 500 curved sheets made of 15 MnNi 63 material. (orig./GL)

  9. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  10. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    Science.gov (United States)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    , for as high as 1600 °C, the non-reactive LM was experimentally confirmed not to show any chemical reaction with air or moisture in the air. This experimental work confirmed the excellent compatibility behaviors between the LM as a PWR fuel gap filler and stainless steel and high-density concrete in the high-temperature regime.

  11. Complementation of a primer binding site-impaired murine leukemia virus-derived retroviral vector by a genetically engineered tRNA-like primer

    DEFF Research Database (Denmark)

    Lund, Anders Henrik; Duch, M; Lovmand, J;

    1997-01-01

    , but not with a noncomplementary tRNA-like molecule. The engineered primer was shown to be involved in both the initiation of first-strand synthesis and second-strand transfer. These results provide an in vivo demonstration that the retroviral replication machinery may recognize sequence complementarity rather than actual primer...... binding site and 3' primer sequences. Use of mutated primer binding site vectors replicating via engineered primers may add additional control features to retroviral gene transfer technology....

  12. PWR Secondary Water Chemistry Control Status: A Summary of Industry Initiatives, Experience and Trends Relative to the EPRI PWR Secondary Water Chemistry Guidelines

    International Nuclear Information System (INIS)

    The latest revision of the EPRI Pressurized Water Reactor (PWR) Secondary Water Chemistry Guidelines was issued in February 2009. The Guidelines continue to focus on minimizing stress corrosion cracking (SCC) of steam generator tubes, as well as minimizing degradation of other major components / subsystems of the secondary system. The Guidelines provide a technically-based framework for a plant-specific and effective PWR secondary water chemistry program. With the issuance of Revision 7 of the Guidelines in 2009, many plants have implemented changes that allow greater flexibility on startup. For example, the previous Guidelines (Revision 6) contained a possible low power hold at 5% power and a possible mid power hold at approximately 30% power based on chemistry constraints. Revision 7 has established a range over which a plant-specific value can be chosen for the possible low power hold (between 5% and 15%) and mid power hold (between 30% and 50%). This has provided plants the ability to establish significant plant evolutions prior to reaching the possible power hold; such as establishing seal steam to the condenser, placing feed pumps in service, or initiating forward flow of heater drains. The application of this flexibility in the industry will be explored. This paper also highlights the major initiatives and industry trends with respect to PWR secondary chemistry; and outlines the recent work to effectively address them. These will be presented in light of recent operating experience, as derived from EPRI's PWR Chemistry Monitoring and Assessment (CMA) program (which contains more than 400 cycles of operating chemistry data). (authors)

  13. Primers for Phylogeny Reconstruction in Bignonieae (Bignoniaceae Using Herbarium Samples

    Directory of Open Access Journals (Sweden)

    Alexandre R. Zuntini

    2013-08-01

    Full Text Available Premise of the study: New primers were developed for Bignonieae to enable phylogenetic studies within this clade using herbarium samples. Methods and Results: Internal primers were designed based on available sequences of the plastid ndhF gene and the rpl32-trnL intergenic spacer region, and the nuclear gene PepC. The resulting primers were used to amplify DNA extracted from herbarium materials. High-quality data were obtained from herbarium samples up to 53 yr old. Conclusions: The standardized methodology allows the inclusion of herbarium materials as alternative sources of DNA for phylogenetic studies in Bignonieae.

  14. A physicists guide to The Los Alamos Primer

    Science.gov (United States)

    Reed, B. Cameron

    2016-11-01

    In April 1943, a group of scientists at the newly established Los Alamos Laboratory were given a series of lectures by Robert Serber on what was then known of the physics and engineering issues involved in developing fission bombs. Serber’s lectures were recorded in a 24 page report titled The Los Alamos Primer, which was subsequently declassified and published in book form. This paper describes the background to the Primer and analyzes the physics contained in its 22 sections. The motivation for this paper is to provide a firm foundation of the background and contents of the Primer for physicists interested in the Manhattan Project and nuclear weapons.

  15. Relativistic Astrophysics and Cosmology: A Primer

    Energy Technology Data Exchange (ETDEWEB)

    Abramowicz, Marek A [Department of Astronomy and Astrophysics, Chalmers University of Technology, 41296 Goeteborg (Sweden)

    2007-10-21

    'Relativistic Astrophysics and Cosmology: A Primer' by Peter Hoyng, was published last year by Springer. The book is based on lectures given by the author at University of Utrecht to advanced undergraduates. This is a short and scholarly book. In about 300 pages, the author has covered the most interesting and important applications of Albert Einstein's general relativity in present-day astrophysics and cosmology: black holes, neutron stars, gravitational waves, and the cosmic microwave background. The book stresses theory, but also discusses several experimental and observational topics, such as the Gravity Probe B mission, interferometer detectors of gravitational waves and the power spectrum of the cosmic microwave background. The coverage is not uniform. Some topics are discussed in depth, others are only briefly mentioned. The book obviously reflects the author's own research interests and his preferences for specific mathematical methods, and the choice of the original artwork that illustrates the book (and appears on its cover) is a very personal one. I consider this personal touch an advantage, even if I do not always agree with the author's choices. For example, I employ Killing vectors as a very useful mathematical tool not only in my research on black holes, but also in my classes. I find that my students prefer it when discussions of particle, photon and fluid motion in the Schwarzschild and Kerr spacetimes are based explicitly and directly on the Killing vectors rather than on coordinate calculations. The latter approach is, of course, the traditional one, and is used in Peter Hoyng's book. Reading the book is a stimulating experience, because the reader can almost feel the author's presence. The author's opinions, his mathematical taste, his research pleasures, and his pedagogical passion are apparent everywhere. Lecturers contemplating a new course on relativistic astrophysics could adopt Hoyng's book as

  16. propósito del primer centenario

    Directory of Open Access Journals (Sweden)

    René Martínez Lemoine

    2007-01-01

    Full Text Available Este trabajo, trata de las apreciaciones públicas y urbanas que se dieron en la ciudad de Santiago a instancias de la celebración del primer Centenario de la República, en el año 1910. Es una rememoranza de aquellas visiones que la ciudad sustrajo a los medios periodísticos, personajes y liderazgos de una sociedad autosatisfecha por el progreso notable alcanzado por el país en las últimas décadas del siglo XIX y los primeros años del siglo XX, pero que ignoró los movimientos obreros y expresiones de disconformidad social que ya comenzaban a ebullir en esos años. La ciudad, como una vasta, compleja y heterogénea construcción en el espacio, erigida a través de las edades por innumerables y, la más de las veces, anónimos constructores, representa la mayor suma de obra humana acumulada en el tiempo, en la que cada generación va dejando una muestra de su aporte en vivienda, espacios, instalaciones y monumentos, vale decir, de su particular cultura y modo de vida en su propio tiempo. Ciertamente, cada ciudad es historia y memoria de sí misma, testimonio permanente de la continuidad del hombre y de la sociedad humana con su propio pasado. En ese sentido, como somos herederos de nuestra historia y de los hombres y mujeres que construyeron y legaron las ciudades en las que vivimos, es importante rescatar esos valores culturales, sociales, arquitectónicos y urbanísticos de modo de visualizar el paso del tiempo, que se materializa, se hace objeto y se torna visible en la ciudad, en la medida que nos “cuenta” algo.

  17. Steam Generator Chemical Cleaning Application: Korean Experience in PWR NPP

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power (KHNP) performed an EPRI/SGOG chemical cleaning of the secondary side of the steam generators at Ulchin Unit 3 (UCN3) in March 2011 and at Ulchin Unit 4 (UCN4) in September 2011. The steam generator chemical cleaning (SGCC) was performed with venting at the top-of-tube sheet (TTS) and at tube support plates (TSPs) 4, 5, 6, 7, 8, 9, and 10. A primary objective of this SGCC was to address outer diameter stress corrosion cracking (ODSCC), which has been observed at the TTS and TSPs in the UCN3 SGs. The EPRI/SGOG process has been shown to effectively reduce prevailing ODSCC rates at the TTS and TSPs, particularly when applied with periodic venting in this application. This was the first full-length SGCC campaign with venting performed in Korea. Ulchin Unit 3 commenced commercial operation in August 1998 and Ulchin Unit 4 commenced commercial operation in December 1999. UCN3 and UCN4 are a two-loop pressurized water reactor (PWR) of the Korea Standard Nuclear Plant (KSNP) design. The SGs contain high-temperature mill annealed (HTMA) Alloy 600 tubing and are similar in design to the Combustion Engineering CE-80. The KSNP SGs have been susceptible to outer diameter stress corrosion cracking (ODSCC), which is consistent with operating experience for other SGs containing Alloy 600HTMA tubing material. The UCN3/4 SGs have recently begun to experience ODSCC. Hankook Jungsoo Industries Co., Ltd (HaJI) was selected as the cleaning vendor by KHNP. To date, HaJI has completed five Advanced Scale Conditioning Agent (ASCA) cleaning applications and two EPRI/SGOG Steam Generator Chemical Cleaning (SGCC) campaigns for KHNP. The goal of total deposit removal of the applications were successfully achieved and the amounts are 3,579 kg at UCN3 and 3,786 kg at UCN4 which values were estimated before each cleaning by analysing ECT signal and liquid samples from the SGs. The deposits from the SGs were primarily composed of magnetite. There were no chemical

  18. An investigation into the efficiency of ion-exchange membranes in simulated PWR coolants

    International Nuclear Information System (INIS)

    This report describes an investigation of the retention efficiency of cation-exchange membranes for magnesium, calcium and nickel ions in PWR-coolant type solutions containing 2 ppm lithium (as lithium hydroxide) and 1000 ppm boron (as boric acid). By analysis of the membranes themselves or of the effluent, the retention characteristics of the membranes in various experimental conditions have been examined. (author)

  19. Radiation embrittlement of Sizewell 'B' PWR pressure vessel during a 40 year lifetime

    International Nuclear Information System (INIS)

    Irradiation embrittlement data produced from PWR surveillance and materials test reactor accelerated experiments on low alloy Mn Mo Ni pressure vessel steels, which satisfy the Sizewell 'B' forging materials specification on copper (<= 0.09% wt.%) and nickel (<= 0.85 wt.%), have been examined. (author)

  20. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  1. Response of pressurized water reactor (PWR) to network power generation demands

    International Nuclear Information System (INIS)

    The flexibility of the PWR type reactor in terms of response to the variations of the network power demands, is demonstrated. The factors that affect the transitory flexibility and some design prospects that allow the reactor fits the requirements of the network power demands, are also discussed. (M.J.A.)

  2. PWR nozzle 'crotch corner' inspection: an effective additional ultrasonic technique for radial cracks

    International Nuclear Information System (INIS)

    An ultrasonic non-destructive technique for testing the integrity of the nozzle crotch corner of a PWR pressure vessel is described which uses two angled probes to detect the specular reflection from one probe to the other via a crack lying in the important radial plane of the nozzle. (U.K.)

  3. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    International Nuclear Information System (INIS)

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO2 fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  4. Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed

    International Nuclear Information System (INIS)

    In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores

  5. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  6. Application of the BEACON-TSM system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    BEACON-TSM is an advanced system of the operation support of PWR reactors that combines the capabilities of an advanced nodal neutronic model and the measures of the instrumentation available in plant to determine, accurately and continuously, the distribution of power in the core and the available margins to the limits of the beak factors.

  7. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    International Nuclear Information System (INIS)

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.)

  8. ''The place of the fatigue risks in the PWR maintenance programs''

    International Nuclear Information System (INIS)

    The parts of components submitted to fatigue risk are more particularly controlled in operation. Three main cases are identified: the mechanical oligo-cyclic fatigue, the vibrating fatigue and the thermal fatigue. These cases are presented in this paper. As a precaution a complementary investigation program is implementing during the Number two decennial inspections of the 900 MW PWR. (A.L.B.)

  9. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    International Nuclear Information System (INIS)

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables

  10. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  11. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

  12. The sorption of iodine by an inorganic zinc primer

    Energy Technology Data Exchange (ETDEWEB)

    Evans, G.J.; Bekeris, P.A. [Toronto Univ., ON (Canada). Dept. of Chemical Engineering and Applied Chemistry

    1996-12-01

    The purpose of this work was to identify and evaluate significant parameters in the sorption of I{sub 2}(g) onto Carbo Zinc 11 inorganic primer, a paint used in the containment structure of some CANDU reactors. Air containing known amounts of {sup 131}I{sub 2}(g) was passed through 0.64 cm diameter glass tubing coated on the inner surface with paint. The accumulation of iodine on the surface was continuously monitored using two scintillation detectors. The test parameters covered were relative humidity, flow rate, I{sub 2} concentration and paint temperature. Adsorption was rapid at 23{sup o}C and predominantly gas phase mass transfer limited: the deposition velocity of 0.7{+-}0.4 cm/s was similar to the gas phase mass transfer coefficient of 1.2 cm/s estimated for the system. The deposition velocity observed at a higher paint surface temperature was an order of magnitude smaller. A similar deposition velocity was observed at 23{sup o}C for adsorption of I{sub 2}(g) from essentially dry air suggesting that the low deposition velocity observed for high surface temperature was limited by the amount of water on the paint surface. The rate of adsorption on the paint was directly proportional to the I{sub 2}(g) concentration over the range in concentration studied. The majority of the iodine retained by the paint could not be removed by washing with methanol or chloroform, but it was removed by water indicating that it was in an ionic form. Analysis of the speciation of the iodine in the wash water indicated that only a third of it was in the form of I{sup -}; the form of the remaining iodine could not be resolved. Desorption from the paint was negligible at room temperature but was detectable at higher temperatures. These low desorption rates and the ionic nature of the surface iodine indicated that adsorption occurred predominantly through a chemisorption process. A number of possible mechanisms were proposed. (author) 5 figs., 2 tabs., 6 refs.

  13. Assessment of void swelling in austenitic stainless steel PWR core internals.

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling

  14. PrecisePrimer: an easy-to-use web server for designing PCR primers for DNA library cloning and DNA shuffling.

    Science.gov (United States)

    Pauthenier, Cyrille; Faulon, Jean-Loup

    2014-07-01

    PrecisePrimer is a web-based primer design software made to assist experimentalists in any repetitive primer design task such as preparing, cloning and shuffling DNA libraries. Unlike other popular primer design tools, it is conceived to generate primer libraries with popular PCR polymerase buffers proposed as pre-set options. PrecisePrimer is also meant to design primers in batches, such as for DNA libraries creation of DNA shuffling experiments and to have the simplest interface possible. It integrates the most up-to-date melting temperature algorithms validated with experimental data, and cross validated with other computational tools. We generated a library of primers for the extraction and cloning of 61 genes from yeast DNA genomic extract using default parameters. All primer pairs efficiently amplified their target without any optimization of the PCR conditions. PMID:24829457

  15. New universal matK primers for DNA barcoding angiosperms

    Institute of Scientific and Technical Information of China (English)

    Jing YU; Jian-Hua XUE; Shi-Liang ZHOU

    2011-01-01

    The chloroplast maturase K gene (matK) is one of the most variable coding genes of angiosperms and has been suggested to be a "barcode" for land plants. However, matK exhibits low amplification and sequencing rates due to low universality of currently available primers and mononucleotide repeats. To resolve these technical problems, we evaluated the entire matK region to find a region of 600-800 bp that is highly variable, represents the best of all matK regions with priming sites conservative enough to design universal primers, and avoids the mononucleotide repeats. After careful evaluation, a region in the middle was chosen and a pair of primers named natK472F and matK1248R was designed to amplify and sequence the matK fragment of approximately 776 bp. This region encompasses the most variable sites, represents the entire matK region best, and also exhibits high amplification rates and quality of sequences. The universality of this primer pair was tested using 58 species from 47 families of angiosperm plants. The primers showed a strong amplification (93.1%) and sequencing (92.6%)successes in the species tested. We propose that the new primers will solve, in part, the problems encountered when using matK and promote the adoption of matK as a DNA barcode for angiosperms.

  16. Analysis of difficulties accounting and evaluating nuclear material of PWR fuel plant

    International Nuclear Information System (INIS)

    Background: Nuclear materials accountancy must be developed for nuclear facilities, which is required by regulatory in China. Currently, there are some unresolved problems for nuclear materials accountancy of bulk nuclear facilities. Purpose: The retention values and measurement errors are analyzed in nuclear materials accountancy of Power Water Reactor (PWR) fuel plant to meet the regulatory requirements. Methods: On the basis of nuclear material accounting and evaluation data of PWR fuel plant, a deep analysis research including ratio among random error variance, long-term systematic error variance, short-term systematic error variance and total error involving Material Unaccounted For (MUF) evaluation is developed by the retention value measure in equipment and pipeline. Results: In the equipment pipeline, the holdup estimation error and its total proportion are not more than 5% and 1.5%, respectively. And the holdup estimation can be regraded as a constant in the PWR nuclear material accountancy. Random error variance, long-term systematic error variance, short-term systematic error variance of overall measurement, and analytical and sampling methods are also obtained. A valuable reference is provided for nuclear material accountancy. Conclusion: In nuclear material accountancy, the retention value can be considered as a constant. The long-term systematic error is a main factor in all errors, especially in overall measurement error and sampling error: The long-term systematic errors of overall measurement and sampling are considered important in the PWR nuclear material accountancy. The proposals and measures are applied to the nuclear materials accountancy of PWR fuel plant, and the capacity of nuclear materials accountancy is improved. (authors)

  17. Methodology and software tools used by IPSN crisis centre experts during an emergency in a French PWR

    International Nuclear Information System (INIS)

    The French nuclear power plants presently in operation are standard pressurized water reactors. Due to the potential consequences of an accident in this type of installation, a national emergency organization was constituted which has the capacity to implement countermeasures for the control of the risks for the surrounding population. The Institute for Nuclear Safety and Protection (IPSN), the technical support of the French nuclear safety authority, has thus defined and constituted a support system which could, in case of an emergency occurring in a French PWR, help to reach this aim. First of all, IPSN has defined and utilizes in its emergency technical centre a methodology to evaluate the plant status and to estimate the evolution of the accident, in order to be able to calculate the consequences of releases in the mean terms. To apply this methodology, a support system, constituted by software tools, was developed and is used in such case. To help decision, two systems are currently used: SESAME for the evaluation of the installation status and potential releases and CONRAD for radiological consequences calculations. The diagnosis of the status of a PWR during an accident is based on the analysis of plant-specific data. The information transmitted from the plant is organized in such way that the expert team assesses rapidly the status of the different safety functions and barriers. The specific tools of the SESAME system are used to quantify parameters such as break size or potential fission product release within or outside the plant. The prognosis of the evolution of safety functions and barriers is based on the assessment of the current and future availability of the safety systems and on extrapolations to forecast the evolution of the accident (time to core uncover and core degradation, fission product release prognosis etc.). The system CONRAD is mainly oriented to the prediction of the consequences in the early phase of the accident for the short term

  18. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  19. TSUNAMI Primer: A Primer for Sensitivity/Uncertainty Calculations with SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Rearden, Bradley T [ORNL; Mueller, Don [ORNL; Bowman, Stephen M [ORNL; Busch, Robert D. [University of New Mexico, Albuquerque; Emerson, Scott [University of New Mexico, Albuquerque

    2009-01-01

    This primer presents examples in the application of the SCALE/TSUNAMI tools to generate k{sub eff} sensitivity data for one- and three-dimensional models using TSUNAMI-1D and -3D and to examine uncertainties in the computed k{sub eff} values due to uncertainties in the cross-section data used in their calculation. The proper use of unit cell data and need for confirming the appropriate selection of input parameters through direct perturbations are described. The uses of sensitivity and uncertainty data to identify and rank potential sources of computational bias in an application system and TSUNAMI tools for assessment of system similarity using sensitivity and uncertainty criteria are demonstrated. Uses of these criteria in trending analyses to assess computational biases, bias uncertainties, and gap analyses are also described. Additionally, an application of the data adjustment tool TSURFER is provided, including identification of specific details of sources of computational bias.

  20. Design Development and Verification of a System Integrated Modular PWR

    International Nuclear Information System (INIS)

    An advanced PWR with a rated thermal power of 330 MW has been developed at the Korea Atomic Energy Research Institute (KAERI) for a dual purpose: seawater desalination and electricity generation. The conceptual design of SMART ( System-Integrated Modular Advanced ReacTor) with a desalination system was already completed in March of 1999. The basic design for the integrated nuclear desalination system is currently underway and will be finished by March of 2002. The SMART co-generation plant with the MED seawater desalination process is designed to supply forty thousand (40,000) tons of fresh water per day and ninety (90) MW of electricity to an area with approximately a ten thousand (100,000) population or an industrialized complex. This paper describes advanced design features adopted in the SMART design and also introduces the design and engineering verification program. In the beginning stage of the SMART development, top-level requirements for safety and economics were imposed for the SMART design features. To meet the requirements, highly advanced design features enhancing the safety, reliability, performance, and operability are introduced in the SMART design. The SMART consists of proven KOFA (Korea Optimized Fuel Assembly), helical once-through steam generators, a self-controlled pressurizer, control element drive mechanisms, and main coolant pumps in a single pressure vessel. In order to enhance safety characteristics, innovative design features adopted in the SMART system are low core power density, large negative Moderator Temperature Coefficient (MTC), high natural circulation capability and integral arrangement to eliminate large break loss of coolant accident, etc. The progression of emergency situations into accidents is prevented with a number of advanced engineered safety features such as passive residual heat removal system, passive emergency core cooling system, safeguard vessel, and passive containment over-pressure protection. The preliminary

  1. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  2. Development and application of methods and computer codes of fuel management and nuclear design of reload cycles in PWR

    International Nuclear Information System (INIS)

    Description of methods and computer codes for Fuel Management and Nuclear Design of Reload Cycles in PWR, developed at JEN by adaptation of previous codes (LEOPARD, NUTRIX, CITATION, FUELCOST) and implementation of original codes (TEMP, SOTHIS, CICLON, NUDO, MELON, ROLLO, LIBRA, PENELOPE) and their application to the project of Management and Design of Reload Cycles of a 510 Mwt PWR, including comparison with results of experimental operation and other calculations for validation of methods. (author)

  3. Studies on the Primers Screening for AFLP Fingerprints of Rice Cultivars

    Institute of Scientific and Technical Information of China (English)

    ZHANG Chun-qing; JIA Ji-zeng

    2002-01-01

    AFLP(amplified fragment length polymorphism) is a very powerful fingerprinting technology.The key of making variety fingerprints is to select specific powerful primers for each crop. A quick and effective procedure for selecting AFLP primers for rice variety fingerprinting was established as the following: (1)Choose 3 or more group materials that have close genetic relations. (2) Select potential polymorphic primers from primer pairs that are 2 + 2 primer crosses and same at two ends. (3) Recombine the selected potential polymorphic primers and choosing more polymorphic primers. (4) Add one selecting base at one end to become 2 + 3 or 3 + 2primers and further selecting more polymorphic primers. Some primers were selected with this procedure, such as M21Ps7 and M73P17, with which the fingerprints had more polymorphism and high quality.

  4. Friction Stir Welding of Shipbuilding Steel with Primer

    Directory of Open Access Journals (Sweden)

    José Azevedo

    2016-03-01

    Full Text Available Abstract Friction Stir Welding has proven its merits for welding of aluminium alloys and is focused in expanding its material database to steel and titanium and also to assess new joint configurations. The use of welded structures in shipbuilding industry has a long tradition and continuously seeks for innovation in terms of materials and processes maintaining, or even, reducing costs. Several studies have been performed in the past years on FSW of steel. However, just recently were reported defect-free welds, free of martensite with stable parameters in steel without Primer. FSW of steel with primer has not been addressed. This work aims to fulfil a knowledge gap related to the use of friction stir for welding shipbuilding steel by analysing the effect of welding parameters on the metallurgical characteristics and mechanical properties of welds obtained with an innovative FSW tool in joining steel plates with a primer. Welds were performed in 4mm thick GL-A36 steel plates painted with a zinc based primer followed by a detailed microscopic, chemical and mechanical analysis. The results that matching fatigue properties are obtained using this technique, in FSW of shipbuilding steel with Primer.

  5. Factores asociados a la recidiva de un neumotórax espontáneo primer episodio. Estudio de casos y controles.

    OpenAIRE

    Raúl Oyarce L.; Egidio Céspedes G.; Felipe Fernández B.; Julio Correa S.

    2011-01-01

    INTRODUCCIÓN: El neumotórax espontáneo (NEP) es la presencia de aire en la pleura sin etiología identificable. Después de un primer episodio 30% recidiva, aún sin consenso del manejo de ambos. Postulamos relación entre datos del paciente consultante por NEP con la recidiva de un neumotórax espontáneo primario primer episodio (NEPPE). OBJETIVO: Identificar características del paciente recidivado del NEPPE (RNEP) versus el paciente sin recidiva. MATERIAL Y MÉTODO: Estudio de casos y controles. ...

  6. Techniques and devices developed by the CEA for hot cell and in-situ examinations of PWR components and PWR fuel assembliess after irradiation

    International Nuclear Information System (INIS)

    Within the framework of the electro-nuclear development of the PWR system, the CEA has provided itself with facilities for developing techniques for analyzing assemblies, pins and fuels. These are examinations and tests on irradiated heads and assemblies with the aid of the Fuel Examination Module (FEM), of machining of assemblies and examinations in the Celimene hot laboratory or detailed examinations and analyses on fuel elements using eddy currents, the electronic microprobe and the Fisher ''permeascope'' which enables the outline of the oxide coat present on the cladding to be followed

  7. MCMC-ODPR: Primer design optimization using Markov Chain Monte Carlo sampling

    Directory of Open Access Journals (Sweden)

    Kitchen James L

    2012-11-01

    Full Text Available Abstract Background Next generation sequencing technologies often require numerous primer designs that require good target coverage that can be financially costly. We aimed to develop a system that would implement primer reuse to design degenerate primers that could be designed around SNPs, thus find the fewest necessary primers and the lowest cost whilst maintaining an acceptable coverage and provide a cost effective solution. We have implemented Metropolis-Hastings Markov Chain Monte Carlo for optimizing primer reuse. We call it the Markov Chain Monte Carlo Optimized Degenerate Primer Reuse (MCMC-ODPR algorithm. Results After repeating the program 1020 times to assess the variance, an average of 17.14% fewer primers were found to be necessary using MCMC-ODPR for an equivalent coverage without implementing primer reuse. The algorithm was able to reuse primers up to five times. We compared MCMC-ODPR with single sequence primer design programs Primer3 and Primer-BLAST and achieved a lower primer cost per amplicon base covered of 0.21 and 0.19 and 0.18 primer nucleotides on three separate gene sequences, respectively. With multiple sequences, MCMC-ODPR achieved a lower cost per base covered of 0.19 than programs BatchPrimer3 and PAMPS, which achieved 0.25 and 0.64 primer nucleotides, respectively. Conclusions MCMC-ODPR is a useful tool for designing primers at various melting temperatures at good target coverage. By combining degeneracy with optimal primer reuse the user may increase coverage of sequences amplified by the designed primers at significantly lower costs. Our analyses showed that overall MCMC-ODPR outperformed the other primer-design programs in our study in terms of cost per covered base.

  8. Mercury in the environment : a primer

    Energy Technology Data Exchange (ETDEWEB)

    Lourie, B.; Glenn, W. (ed.); Ogilvie, K.; Everhardus, E.; Friesen, K.; Rae, S.

    2003-06-01

    This report provides an overview of the occurrence and effects of mercury in the environment and its impacts on human health. Low levels of mercury occur naturally everywhere in the environment in plants, animals, rocks and air. Incidental emissions occur when natural mercury is released to the environment through human activity. In Canada, coal burning and metal processing are the two largest point sources of atmospheric mercury emissions. Energy facilities have the option to invest in expensive control technologies for coal plants, or they can generate electricity from alternative energy sources. Energy conservation, however, offers the greatest overall benefits for the environment and the public. Mercury can also be released when products containing mercury (such as electrical switches, thermostats, dental amalgam, and thermometers) are broken while in use, or when they are crushed in garbage trucks and dumped in landfills. Source separation is the best way to reduce waste-related emissions. Once mercury is released to the natural environment, it can be transported long distances through air or watercourses. It is volatile, therefore evaporates readily to the atmosphere where it may do one of three things: it may fall out near the point where it was emitted; it may be transported long distances to some point downwind; or, it may enter the global atmospheric mercury pool where it will circle the globe for a year or more within the Earth's major weather systems before being deposited. Data from Canada's National Pollutant Release Inventory indicates that mercury releases and transfers total 28,674 kg per year. The most critical component of the mercury cycle is the conversion of inorganic forms of mercury to the organic compound methylmercury which is more toxic to humans. Most concern about mercury focuses on lakes and other aquatic ecosystems. Fish in hydroelectric reservoirs have been found to contain elevated methylmercury levels because natural

  9. KENO-VI Primer: A Primer for Criticality Calculations with SCALE/KENO-VI Using GeeWiz

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, Stephen M [ORNL

    2008-09-01

    The SCALE (Standardized Computer Analyses for Licensing Evaluation) computer software system developed at Oak Ridge National Laboratory is widely used and accepted around the world for criticality safety analyses. The well-known KENO-VI three-dimensional Monte Carlo criticality computer code is one of the primary criticality safety analysis tools in SCALE. The KENO-VI primer is designed to help a new user understand and use the SCALE/KENO-VI Monte Carlo code for nuclear criticality safety analyses. It assumes that the user has a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with SCALE/KENO-VI in particular. The primer is designed to teach by example, with each example illustrating two or three features of SCALE/KENO-VI that are useful in criticality analyses. The primer is based on SCALE 6, which includes the Graphically Enhanced Editing Wizard (GeeWiz) Windows user interface. Each example uses GeeWiz to provide the framework for preparing input data and viewing output results. Starting with a Quickstart section, the primer gives an overview of the basic requirements for SCALE/KENO-VI input and allows the user to quickly run a simple criticality problem with SCALE/KENO-VI. The sections that follow Quickstart include a list of basic objectives at the beginning that identifies the goal of the section and the individual SCALE/KENO-VI features that are covered in detail in the sample problems in that section. Upon completion of the primer, a new user should be comfortable using GeeWiz to set up criticality problems in SCALE/KENO-VI. The primer provides a starting point for the criticality safety analyst who uses SCALE/KENO-VI. Complete descriptions are provided in the SCALE/KENO-VI manual. Although the primer is self-contained, it is intended as a companion volume to the SCALE/KENO-VI documentation. (The SCALE manual is provided on the SCALE installation DVD.) The primer provides specific examples of

  10. An Empirical Approach to Bounding the Axial Reactivity Effects of PWR Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    P. M. O' Leary; J. M. Scaglione

    2001-04-04

    One of the significant issues yet to be resolved for using burnup credit (BUC) for spent nuclear fuel (SNF) is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters (such as local power, fuel temperature, moderator temperature, burnable poison rod history, and soluble boron concentration) affect the isotopic inventory of fuel that is depleted in a pressurized water reactor (PWR). However, obtaining the detailed operating histories needed to model all PWR fuel assemblies to which BUC would be applied is an onerous and costly task. Simplifications therefore have been suggested that could lead to using ''bounding'' depletion parameters that could be broadly applied to different fuel assemblies. This paper presents a method for determining a set of bounding depletion parameters for use in criticality analyses for SNF.

  11. RANS modeling for flow in nuclear fuel bundle in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    This paper presents use of Reynolds-Averaged Navier-Stokes (RANS) based turbulence model for single-phase CFD analysis of flow in Pressurized Water Reactor (PWR) Assemblies. An open source code called OpenFoam was used for computational fluid dynamics (CFD) study using computational meshes generated using Shari Harpoon. The PWR assembly design used in this analysis represents a 5 x 5 pin design including structural grid equipped with mixing vanes. The design specifications used in this study were obtained from the experimental setup at Texas A&M University and the results obtained are used to validate the CFD software, algorithm, and the turbulence model used in this analysis. (author)

  12. Study on severe accident mitigation measures for the development of PWR SAMG

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

  13. Improving electron beam weldability of heavy steel plate for PWR-steam generator

    International Nuclear Information System (INIS)

    There are plans to install or replace many PWR-steam generators. Electron beam welding (EBW) can greatly reduce the welding period compared to conventional welding methods (narrow-gap GMAW and SAW). The problems in applying EBW are to prevent weld defects and to improve the toughness of the weld metal. Successful defect-free welding procedures were established for thick steel plates. As for weld metal toughness, we investigated the factors that deteriorate weld metal toughness of EBW and made clear the manufacturing process which utilizes a new secondary refining process and a high-torque mill in actual mass-production. As a result, application of EBW to PWR-steam generators has become possible and large amounts of ASTM A533B C12 (JIS SQV2B) steel plates for EBW have come to be produced. We evaluated EBW base metal and weld joints including fracture toughness. (author). 12 refs., 8 figs., 6 tabs

  14. Optimization of small long-life PWR based on thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subkhi, Moh Nurul, E-mail: nsubkhi@students.itb.ac.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung (Indonesia); Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung Djati Bandung Jalan A.H Nasution 105 Bandung (Indonesia); Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Waris, Abdul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {sup 231}Pa give low excess reactivity.

  15. Conceptual design study of small long-life PWR based on thorium cycle fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subkhi, M. Nurul [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) and Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung (Indonesia); Su' ud, Zaki; Waris, Abdul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung) (Indonesia)

    2014-09-30

    A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation.

  16. Review of the high conversion-type core study. Review about PWR

    International Nuclear Information System (INIS)

    The study of the high conversion light water reactor was proposed by Edlund in the United States of America in 1975. This theme was aggressively studied in the 1980s. As the reason, the increase of Pu produced from the reprocessing and the delay of practical use of FBR is given. A high converter core can realize comparatively easily by changing core of existing PWR to the tight and short core. In this report, the high converter core study about PWR were reviewed. In the United States of America, the study has already ended but an aggressive study is carried forward in Germany, Japan and so on. In addition to the reactor physics computation, the wide range study of such as critical experiment, conduct experiment of the heat transfer, the fracture behavior of fuel and the reactor-type strategy are carried forward. To investigate these studies is extremely useful in examining a future. (author)

  17. Crud formation on low duty PWR fuel in the Halden reactor

    International Nuclear Information System (INIS)

    A previous paper summarised observations on the effects of water chemistry and thermal-hydraulic conditions on crud formation on PWR fuel in the Halden reactor. These observations led to the conclusion that a critical degree of fuel duty (which can be expressed as degree of coolant sub-cooled boiling, void fraction or mass evaporation rate) was required for the formation of tenacious crud deposits. Recent measurements of the oxide layers on low duty PWR fuel have revealed the formation of tenacious crud deposits. This paper describes the operating history of the fuel rods, including water chemistry and thermal-hydraulic conditions, and suggests reasons for the sudden appearance of the crud deposits. (author)

  18. Estimation of the hydrodynamic effects of a LOCA in A 4-loop PWR

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Potapov, S. [Electricite de France (EDF-SEP / AMV), 92 - Clamart (France); Tephany, F. [Electricite de France (EDF SEPTEN), 69 - Villeurbanne (France)

    2001-07-01

    The PWR safety studies involve an analysis of the consequences of a hypothetical rupture of a primary pipe. From the opening tune, the blowdown at the break causes the propagation of an acoustic wave through the whole primary circuit, as well as pipe whipping. The local pressure gaps due to the depressurization wave propagation may induce component recoils and internal structure movements. In parallel with the acoustic wave propagation, the circuit empties progressively first with a monophasic regime and later with a diphasic one. This paper presents a hydrodynamic simulation of the flows in the primary circuit of 4-loop PWR during a LOCA. The results concern the propagation of the depressurization acoustic wave along the circuit, coupled with the transient fluid flows. (authors)

  19. EPRI PWR Safety and Relief Value Test Program: safety and relief valve test report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A safety and relief valve test program was conducted by EPRI for a group of participating PWR utilities to respond to the USNRC recommendations documented in NUREG 0578 Section 2.1.2, and as clarified in NUREG 0737 Item II.D.1.A. Seventeen safety and relief valves representative of those utilized in or planned for use in participating domestic PWR's were tested under the full range of selected test conditions. This report contains a listing of the selected test valves and the corresponding as tested test matrices, valve performance data and principal observations for the tested safety and relief valves. The information contained in this report may be used by the participating utilities in developing their response to the above mentioned USNRC recommendations.

  20. Dosimetry experiment 'Dompac'. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damage characterization

    International Nuclear Information System (INIS)

    For suitable extrapolation of irradiated PWR ferritic steel results, proper irradiation of the pressure vessel has been 'simulated' in test reactor. For this purpose, a huge steel block (20 cm in depth) was loaded with Saclay's graphite (GAMIN) and tungsten damage detectors. Core-block water gap was optimized through spectrum indexes method, by ANISN and SABINE codes so that spectrum in 1/4 thickness matches with ANISN computations for PWR Fessenheim 1. A good experimental agreement is found with calculated dpa damage gradient. 3D Monte Carlo computation (TRIPOLI) was performed on the DOMPAC device, and spectrum indexes evolution was found consistent with experimental results. Surveillance rigs behind a 'thermal shield' were also simulated, including damage and activation monitors. Dosimetry results give an order of magnitude of accuracies involved in projecting steel sample embrittlement to the pressure vessel

  1. Aerosol removal by emergency spray in PWR containment: synthesis of the TOSQAN aerosol tests

    International Nuclear Information System (INIS)

    During the course of a severe accident in a nuclear Pressurized Water Reactor (PWR), containment reactor is pressurized by steam and hydrogen released from a primary circuit breach and distributed into the containment according to convective flows and steam wall condensation. In addition, core degradation leads to fission product release into the containment. Water spraying is used in the containment as mitigation means in order to reduce pressure, to remove fission products and to enhance the gas mixing in case of presence of hydrogen. This paper presents the synthesis of the results of the TOSQAN aerosol program undertaken by the Institut de Radioprotection et de Surete Nucleaire (IRSN) devoted to study the aerosol removal by a spray, for typical accidental thermal hydraulic conditions in PWR containment. (author)

  2. New cyt b gene universal primer set for forensic analysis.

    Science.gov (United States)

    Lopez-Oceja, A; Gamarra, D; Borragan, S; Jiménez-Moreno, S; de Pancorbo, M M

    2016-07-01

    Analysis of mitochondrial DNA, and in particular the cytochrome b gene (cyt b), has become an essential tool for species identification in routine forensic practice. In cases of degraded samples, where the DNA is fractionated, universal primers that are highly efficient for the amplification of the target region are necessary. Therefore, in the present study a new universal cyt b primer set with high species identification capabilities, even in samples with highly degraded DNA, has been developed. In order to achieve this objective, the primers were designed following the alignment of complete sequences of the cyt b from 751 species from the Class of Mammalia listed in GenBank. A highly variable region of 148bp flanked by highly conserved sequences was chosen for placing the primers. The effectiveness of the new pair of primers was examined in 63 animal species belonging to 38 Families from 14 Orders and 5 Classes (Mammalia, Aves, Reptilia, Actinopterygii, and Malacostraca). Species determination was possible in all cases, which shows that the fragment analyzed provided a high capability for species identification. Furthermore, to ensure the efficiency of the 148bp fragment, the intraspecific variability was analyzed by calculating the concordance between individuals with the BLAST tool from the NCBI (National Center for Biotechnological Information). The intraspecific concordance levels were superior to 97% in all species. Likewise, the phylogenetic information from the selected fragment was confirmed by obtaining the phylogenetic tree from the sequences of the species analyzed. Evidence of the high power of phylogenetic discrimination of the analyzed fragment of the cyt b was obtained, as 93.75% of the species were grouped within their corresponding Orders. Finally, the analysis of 40 degraded samples with small-size DNA fragments showed that the new pair of primers permits identifying the species, even when the DNA is highly degraded as it is very common in

  3. New cyt b gene universal primer set for forensic analysis.

    Science.gov (United States)

    Lopez-Oceja, A; Gamarra, D; Borragan, S; Jiménez-Moreno, S; de Pancorbo, M M

    2016-07-01

    Analysis of mitochondrial DNA, and in particular the cytochrome b gene (cyt b), has become an essential tool for species identification in routine forensic practice. In cases of degraded samples, where the DNA is fractionated, universal primers that are highly efficient for the amplification of the target region are necessary. Therefore, in the present study a new universal cyt b primer set with high species identification capabilities, even in samples with highly degraded DNA, has been developed. In order to achieve this objective, the primers were designed following the alignment of complete sequences of the cyt b from 751 species from the Class of Mammalia listed in GenBank. A highly variable region of 148bp flanked by highly conserved sequences was chosen for placing the primers. The effectiveness of the new pair of primers was examined in 63 animal species belonging to 38 Families from 14 Orders and 5 Classes (Mammalia, Aves, Reptilia, Actinopterygii, and Malacostraca). Species determination was possible in all cases, which shows that the fragment analyzed provided a high capability for species identification. Furthermore, to ensure the efficiency of the 148bp fragment, the intraspecific variability was analyzed by calculating the concordance between individuals with the BLAST tool from the NCBI (National Center for Biotechnological Information). The intraspecific concordance levels were superior to 97% in all species. Likewise, the phylogenetic information from the selected fragment was confirmed by obtaining the phylogenetic tree from the sequences of the species analyzed. Evidence of the high power of phylogenetic discrimination of the analyzed fragment of the cyt b was obtained, as 93.75% of the species were grouped within their corresponding Orders. Finally, the analysis of 40 degraded samples with small-size DNA fragments showed that the new pair of primers permits identifying the species, even when the DNA is highly degraded as it is very common in

  4. EDF/CIDEN - ONECTRA: PWR decontamination; EDF/CIDEN - ONECTRA: assainissement REP

    Energy Technology Data Exchange (ETDEWEB)

    Fayolle, P. [EDFICIDEN, 35-37, rue Louis Guerin - B.P. 21212, 69611 Villeurbanne Cedex (France); Orcel, H. [ONECTRA, ZA les Tomples BP45, 26701 Pierrelatte Cedex (France); Wertz, L. [ONECTRA, Le Britannia, Allee C, 20 Bd Eugene Deruelle, 69432 Lyon Cedex 03 (France)

    2010-07-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  5. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

    International Nuclear Information System (INIS)

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs

  6. Problems of control of WWER-type pressurized water reactors (PWR's)

    International Nuclear Information System (INIS)

    The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

  7. Gamma-ray measurements of spent PWR fuel and determination of residual power

    OpenAIRE

    Jansson, Peter; Håkansson, Ane; Bäcklin, Anders

    1997-01-01

    The method for determining residual thermal power in spent BWR fuel described in ISV-4/97 have been used in an extended study where spent PWR fuel assemblies have been considered. The experimental work has been carried out at the interim storage CLAB. By using the 137Cs radiation it is shown in the present study that it is possible to experimentally determine the residual thermal power within 3%.

  8. Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

    Energy Technology Data Exchange (ETDEWEB)

    Stutzmann, A. [Electricite de France (EDF), 93 - Saint-Denis (France)

    1997-07-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  9. Anti -corrosion Effect of ETA on Materials in Secondary Loop of PWR

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    In the world, over sixty percent of nuclear power plant have used advanced amunes ETA(Ethanolamine) as pH control agent in secondary loop of PWR. There are eighty percent of nuclear powerplants using ETA in USA. The corrosion of materials in steam generator (SG) tube and secondary looppower water reactor have been inhibited, the life of SG and the economics of the plant are increasedbecause of using ETA.

  10. Methods and computer programs for PWR's fuel management: Programs Sothis and Ciclon

    International Nuclear Information System (INIS)

    Methos and computer programs developed at JEN for fuel management in PWR are discussed, including scope of model, procedures for sistematic selection of alternatives to be evaluated, basis of model for neutronic calculation, methods for fuel costs calculation, procedures for equilibrium and trans[tion cycles calculation with Soth[s and Ciclon codes and validation of methods by comparison of results with others of reference (author) '

  11. STUDY OF THE THERMAL STRATIFICATION IN PWR REACTORS AND THE PTS (PRESSURIZED THERMAL SHOCK) PHENOMENON

    OpenAIRE

    ROMERO HAMERS, ADOLFO

    2014-01-01

    In the event of hypothetical accident scenarios in PWR, emergency strategies have to be mapped out, in order to guarantee the reliable removal of decay heat from the reactor core, also in case of component breakdown. One essential passive heat removal mechanism is the reflux condensation cooling mode. This mode can appear for instance during a small break loss-of-coolant-accident (LOCA) or because of loss of residual heat removal (RHR) system during mid loop operation at plant outage after th...

  12. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  13. The impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study, PWR [pressurized-water reactor] during an outage

    International Nuclear Information System (INIS)

    This report is the second in a series of case studies designed to evaluate the magnitude of increase in occupational radiation exposures at commercial US nuclear power plants resulting from small incidents or abnormal events. The event evaluated is fuel cladding failure, which can result in elevated primary coolant activity and increased radiation exposure rates within a plant. For this case study, radiation measurements were made at a pressurized-water reactor (PWR) during a maintenance and refueling outage. The PWR had been operating for 22 months with fuel cladding failure characterized as 105 pin-hole leakers, the equivalent of 0.21% failed fuel. Gamma spectroscopy measurements, radiation exposure rate determinations, thermoluminescent dosimeter (TLD) assessments, and air sample analyses were made in the plant's radwaste, pipe penetration, and containment buildings. Based on the data collected, evaluations indicate that the relative contributions of activation products and fission products to the total exposure rates were constant over the duration of the outage. This constancy is due to the significant contribution from the longer-lived isotopes of cesium (a fission product) and cobalt (an activation product). For this reason, fuel cladding failure events remain as significant to occupational radiation exposure during an outage as during routine operations. As documented in the previous case study (NUREG/CR-4485 Vol. 1), fuel cladding failure events increased radiation exposure rates an estimated 540% at some locations of the plant during routine operations. Consequently, such events can result in significantly greater radiation exposure rates in many areas of the plant during the maintenance and refueling outages than would have been present under normal fuel conditions

  14. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    Science.gov (United States)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  15. PWR ENDF/B-VII Cross-Section Libraries for ORIGEN-ARP

    International Nuclear Information System (INIS)

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  16. RNL NDT studies related to PWR pressure vessel inlet nozzle inspection

    International Nuclear Information System (INIS)

    Non-destructive examinations of the Reactor Pressure Vessel (RPV) of a Pressurized Water Reactor (PWR) play an important role in assuring vessel integrity throughout its operational life. Automated ultrasonic techniques for the detection and sizing of flaws in thick-section seam welds and near-surface regions in a PWR RPV have been under development at RNL for some time. Techniques for the inspection of complex geometry welds and other regions of the vessel are now being assessed and further developed as part of the UK NDT development programme in support of the Sizewell PWR. One objective of this programme is to demonstrate that the range of ultrasonic techniques already shown to be effective for the inspection of seam welds and inlet nozzle corner regions, through exercises such as the Defect Detection Trials, can also be effective for inspection of these other vessel regions. The nozzle-to-vessel welds and nozzle crotch corners associated with the RPV water inlet and outlet nozzles are two such regions being examined in this programme. In this paper, a review is given of the work performed at RNL in the development of a laboratory-based inspection system for inlet nozzle inspection. The main features of the system in its current stage of development are explained. (author)

  17. Improving electron beam weldability of heavy steel plates for PWR-steam generator

    International Nuclear Information System (INIS)

    Installation and replacement of many PWR-steam generators are planned inside and outside Japan. The steel plates for steam generators are heavy in thickness, and increase the number of welding passes and prolong the welding time. Electron beam welding (EBW) can greatly reduce the welding period compared with conventional welding methods (narrow-gap gas metal arc welding (GMAW) and submerged arc welding (SAW)). The problems in applying EBW are to prevent weld defects and to improve the toughness of the weld metal. Defect-free welding procedures were successfully established even in thick steel plates. The factors that deteriorate weld-metal (WM) toughness of EBW were investigated. The manufacturing process, which utilizes a new secondary refining process at steelmaking and a high-torque mill at plate mill in actual mass-production, were established. EBW base metal and WM have better properties including fracture toughness than those of conventional welding processes. As a result, an application of EBW to the fabrication of PWR-steam generators has become possible. Large amounts of ASTM A533 Gr B Cl 2 (JIS SQV2B) steel plates in actual PWR-steam generators have come to be produced (more than 1,500 ton) by applying EBW. (author)

  18. Development of mechanical test techniques for structural components of irradiated PWR fuel assembly

    International Nuclear Information System (INIS)

    An increase of fuel burnup and duration of fuel life remains one of the main methods for a nuclear power engineering enhancement. Properties of structural materials providing corrosion resistance, mechanical strength, and dimensional instability of the components of a fuel assembly (FA) are of great importance for fuel operational reliability in such fuel life cycles. Generally, PWR fuel assemblies consist of a top nozzle, spacer grid, bottom nozzle, and guide/instrumentation tubes. The top and bottom nozzle are fixed to the guide tubes using a screw or bulge method. The spacer grid fixed to the guide/instrumentation tubes using a spot weld or bulge method. To understand the in-reactor performance of PWR FA, several devices and test techniques have been developed for mechanical property tests. Among the structural components of PWR FA, a spacer grid, a hold down spring of a top nozzle and a connecting part of FA were considered. Experimental works were carried out for the unirradiated and irradiated components of advanced nuclear fuel assemblies for KSNPs and Westinghouse type PWRs at IMEF (Irradiated Materials Examination Facility) at KAERI. The developed techniques were verified through a hot cell tests. (author)

  19. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method

  20. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  1. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water

    International Nuclear Information System (INIS)

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. δ phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  2. Secure and effective valve stem sealing in PWR power generating plants

    International Nuclear Information System (INIS)

    The PWR power generating plant combines severe operating conditions with the highest safety requirements, making it one of the most demanding environments for seals. An analysis of the conditions inherent in its operation reveals: an aggressive and radioactive fluid at high temperature and pressure; frequent thermal shocks; and hazards for maintenance personnel in the containment area unless the reactor is shut down. The achievement of today's quality and safety standards owes much to the experience, research and testing carried out by the Electricite de France during its graduation from its first nuclear unit to become the world's most important manager of PWR plants with over 45 now under its control. The number of valves involved in the French nuclear program is in excess of 1,300,000. Knowing what the affect of a leak can be, especially if it necessitates a shutdown of the power station, the need to insure the quality of valve sealing can be appreciated. At the beginning of their nuclear building program, the EdF was finding that valves, representing only 2 percent of the investment in a PWR plant, caused 20% of the unwanted outages and cost 60% of the total of plant maintenance. In this report, the author endeavors to show how this problem was solved by team work and concerted action by the EdF, the valve constructors and seal manufacturer, not forgetting the importance of informing and training the maintenance and repair teams within the power stations themselves

  3. A Novel Burnable Absorber Concept for PWR: BigT (Burnable Absorber-Integrated Guide Thimble)

    Energy Technology Data Exchange (ETDEWEB)

    Yahya, Mohdsyukri; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Chung, Chang Kyu [KEPCO Engineering and Construction Company, Daejeon (Korea, Republic of)

    2014-05-15

    This paper presents the essential BigT design concepts and its lattice neutronic characteristics. Neutronic performance of a newly-proposed BA concept for PWR named BigT is investigated in this study. Preliminary lattice analyses of the BigT absorber-loaded WH 17x17 fuel assembly show a high potential of the concept as it performs relatively well in comparison with commercial burnable absorber technologies, especially in managing reactivity depletion and peaking factor. A sufficiently high control rod worth can still be obtained with the BigT absorbers in place. It is expected that with such performance and design flexibilities, any loading pattern and core management objective, including a soluble boron-free PWR, can potentially be fulfilled with the BigT absorbers. Future study involving full 3D reactor core simulations with the BigT absorbers shall hopefully verify this hypothesis. A new burnable absorber design for Pressurized Water Reactor (PWR) named 'Burnable absorber-Integrated control rod Guide Thimble' (BigT) was recently proposed. Unlike conventional burnable absorber (BA) technologies, the BigT integrates BA materials directly into the guide thimble but still allows insertion of control rod (CR). In addition, the BigT offers a variety of design flexibilities such that any loading pattern and core management objective can potentially be fulfilled.

  4. Application of a burnup verification meter to actinide-only burnup credit for spent PWR fuel

    International Nuclear Information System (INIS)

    A measurement system to verify reactor records for burnup of spent fuel at pressurized water reactors (PWR) has been developed by Sandia National Laboratories and tested at US nuclear utility sites. The system makes use of the Fork detector designed at Los Alamos National Laboratory for the safeguards program of the International Atomic Energy Agency. A single-point measurement of the neutrons and gamma- rays emitted from a PWR assembly is made at the center plane of the assembly while it is partially raised from its rack in the spent fuel pool. The objective of the measurements is to determine the variation in burnup assignments among a group of assemblies, and to identify anomalous assemblies that might adversely affect nuclear criticality safety. The measurements also provide an internal consistency check for reactor records of cooling time and initial enrichment. The burnup verification system has been proposed for qualifying spent fuel assemblies for loading into containers designed using burnup credit techniques. The system is incorporated in the US Department of Energy's.''Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages'' [DOE/RW 19951

  5. Layout of the primary circuit with its components for PWR and BWR

    International Nuclear Information System (INIS)

    The light water-moderated and cooled pressurized water reactors and boiling water reactors constitute the basis of economic utilization of nuclear energy all over the world. Pressurized water reactors up to capacities of 3,800 MWth are those most used for power generation. However, their potential capacities exceed 3,800 MWth, so that already in the near future PWR are conseivable which readily generate 1,500 to 2,000 MWe. The main problem for starting the next generation of PWRs are of safety measure and licensing questions. Interesting applications of the PWRs are nuclear district heating, generation of process steam of desalination plants, steam injection into the ground for oil production or chemical factories. A new generation of natural circulation boiling water reactors with a capacity of 200 to 400 MW will be used for development of small industrial areas or for countries without an integral grid system. The natural circulation boiling water reactor will be subject of a separate lecture. Due to the fact of the majority of the PWR all over the world this lecture will discuss mainly PWR design aspects. (orig./RW)

  6. Air Pollution

    Science.gov (United States)

    Air pollution is a mixture of solid particles and gases in the air. Car emissions, chemicals from factories, dust, pollen and ... Ozone, a gas, is a major part of air pollution in cities. When ozone forms air pollution, ...

  7. FACTORES AMBIENTALES QUE AFECTAN LA EDAD AL PRIMER PARTO Y PRIMER INTERVALO DE PARTOS EN VACAS DEL SISTEMA DOBLE PROPOSITO

    OpenAIRE

    Caty Martínez B; Luz Botero A; Oscar Vergara G

    2009-01-01

    Objetivo. Determinar los factores que influyen en la edad al primer parto (AFC) y primer intervalo de parto (PIDP) en hembras bovinas bajo el sistema de doble propósito, en la finca “El Rodeo”, municipio de Magangué, Bolívar - Colombia. Materiales y métodos. Se analizaron 379 datos provenientes de los registros productivos entre los años 1993 hasta 2002, usando el programa estadístico GLM del Statistical Analysis System, donde se obtuvieron la media y el error estándar de cada fuente de vari...

  8. Secondary water chemistry control practices and results of the Japanese PWR plants

    International Nuclear Information System (INIS)

    In Japan, since the start of the operation of the first PWR plant, Mihama Unit-1 in 1970, 24 PWR plants have been built by 2010, and all of them are in operation. Due to the plant-specific needs of management, and by flexibly incorporating the state-of-the-art insights into the design, the system configurations of the plants vary so many as 15 types. Meanwhile, the geographical feature of Japan makes all the Japanese PWR plants to have condensers cooled by sea water, and all the plants have a common system with a full-flow Condensate Polisher System (CPS). To prevent corrosion, continued improvements of the secondary water chemistry management has been performed like other countries, and one of the major features of the Japanese PWR plants is an enhanced provision for the condenser leakage. The water quality of SG (Steam Generator) has been significantly improved by the provision for the sea water leakage, in combination with other improvements in water chemistry management. Also in Japan, almost all of the treatments of the spent polisher resin and the wastewater are performed within the power plant sites. To facilitate the treatment of the waste water and the regeneration of the spent resins, either ammonia or ETA (Ethanol Amine) is selected as the pH adjustment agent for the secondary system water. Also at the ammonia treatment, high pH accomplishes the inhibition of the piping wall thinning and the lower iron transportation into SGs. In addition, the iron transported into the SG is removed by the chemical conditioning treatment called ASCA (Advanced Scale Conditioning Agent). This provides the effective recovery of the SG heat-transfer performance, and the improved SG support plate BEC (Broached Egg Crate) hole blockage rates. Basically in Japan, the secondary water chemistry management has been improved based on a single basic specification, for the variety of the plant configurations, with the plant-specific investigations and analyses. This paper summarizes

  9. Characteristics of the population employed in primer sector in Turkey

    Directory of Open Access Journals (Sweden)

    Bayar Rüya

    2006-01-01

    Full Text Available Activities related to the production of raw material like agriculture husbandry, forestry, fishery are called as primer activities. Especially people living in rural areas earn their livings on primer activities, mainly agriculture. Rural planning is inevitable for providing rural development which has an important place in all development of a country. And achievement of this planning depends on putting forth the characteristics of the population living in rural areas with its different aspects. Therefore, the requirements will be introduced more clearly and the increase in the welfare levels of the people living in rural areas will have been achieved. To achieve the rural development and progress, in addition to the features like the size of agricultural products, products that are cultivated, activities like husbandry, forestry, hunting, etc. and the qualities of the enterprises in which these activities are carried out, policies applied, capital, market and technology, the characteristics of the population employed in this sector is also of importance. Considering these points, what is aimed in this study is to put forth the characteristics of the population employed in primer sector in Turkey. According to the census results of the year 2000 in Turkey 38% of the population is employed, and 48% of this work is in primer sector.

  10. Criticality calculations with MCNP{sup TM}: A primer

    Energy Technology Data Exchange (ETDEWEB)

    Mendius, P.W. [ed.; Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

    1994-08-01

    The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.

  11. Diversity of internal structures in inhibited epoxy primers

    Directory of Open Access Journals (Sweden)

    Anthony E. Hughes

    2015-10-01

    Full Text Available Computed tomography is making a significant impact in the field of materials science in recent years. In this paper the authors report on advances made in three areas of characterization and also identified where further research needs to be focused. First we report on a new approach to data analysis called “Data Constrained Modelling (DCM” in which compositional tomography can be undertaken rather than adsorption or phase contrast tomography. This is achieved by collecting X-ray CT data at different energies and then combining the datasets to reconstruct 3D compositional tomography. Second, on the application of this approach to inhibited primers typical of those used in the aerospace industry. Aerospace primers are effectively composite materials containing inorganic phases which are bound together with a polymer. Understanding the materials science of these systems requires information over several orders of magnitude in length-scale. In this paper we report on how DCM can be used to extend our understanding at the smaller length scales at the limits of resolution of the technique. The third and final advance is in extending the approach to include 4-dimensional studies. In this case we examine the primer before and after leaching. This process causes changes in the primer which can be both detected and quantified using the above approach.

  12. Detecting Lesch-Nyhan syndrome by solid phase primer extension

    Energy Technology Data Exchange (ETDEWEB)

    Shumaker, J.M.; Caskey, C.T. [Baylor College of Medicine, Houston, TX (United States); Metspalu, A.

    1994-09-01

    A mutation detection method based upon the wild type human HPRT sequence is presented for identification of Lesch Nyhan syndrome. The technique consists of performing a biotinlyated PCR amplification of the region of interest, followed by isolation and purification of single stranded template using magnetic separation. Allele-specific primers are annealed adjacent to the potential mutation site on the template. A terminal fluorescent deoxynucleotide addition is performed with a DNA template-dependent polymerase to distinguish between the mutant and wild-type sequence. The products are purified from unincorporated ddNTPs, eluted and finally analyzed on an ABI 373 to identify the mutation. The length of an extension primer is used as a position signature for mutations. The fidelity of nucleotide incorporation provides an excellent signal-to-noise ratio for the detection of nine HPRT mutations within eight cell lines. This method should detect all types of mutations except for repeated sequences that are longer than the primers. Moreover, the method is being extended to a solid support assay, whereby the extension primers are attached to a two-dimensional glass surface. Following extension, the solid support is analyzed for radioactive incorporation. We have shown the sequence determination of a five base region of a wild-type sequence and two different HPRT mutations. As more dense oligonucleotide arrays are produced, this method could be extended to sequence the complete coding region of HPRT.

  13. Criminal Justice in the United States: A Primer

    OpenAIRE

    James B. Jacobs

    2007-01-01

    In this primer on the U.S. criminal justice system, James B. Jacobs, Warren E. Burger Professor of Law at New York University (NYU) and Director of the Center for Research in Crime and Justice at the NYU School of Law, explains the structure and basic jurisprudence of U.S. criminal law and criminal procedure.

  14. Criticality calculations with MCNP{trademark}: A primer

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A. [New Mexico Univ., Albuquerque, NM (United States)

    1994-06-06

    With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.

  15. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  16. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  17. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lapa, Nelbia da Silva

    2005-10-15

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  18. Resin bonding of metal brackets to glazed zirconia with a porcelain primer

    OpenAIRE

    Lee, Jung-Hwan; Lee, Milim; Kim, Kyoung-Nam; Hwang, Chung-Ju

    2015-01-01

    Objective The aims of this study were to compare the shear bond strength between orthodontic metal brackets and glazed zirconia using different types of primer before applying resin cement and to determine which primer was more effective. Methods Zirconia blocks were milled and embedded in acrylic resin and randomly assigned to one of four groups: nonglazed zirconia with sandblasting and zirconia primer (NZ); glazed zirconia with sandblasting, etching, and zirconia primer (GZ); glazed zirconi...

  19. PriSM: a primer selection and matching tool for amplification and sequencing of viral genomes

    OpenAIRE

    Yu, Qing; Ryan, Elizabeth M; Allen, Todd M.; Birren, Bruce W.; Henn, Matthew R.; Lennon, Niall J.

    2010-01-01

    Summary: PriSM is a set of algorithms designed to select and match degenerate primer pairs for the amplification of viral genomes. The design of panels of hundreds of primer pairs takes just hours using this program, compared with days using a manual approach. PriSM allows for rapid in silico optimization of primers for downstream applications such as sequencing. As a validation, PriSM was used to create an amplification primer panel for human immunodeficiency virus (HIV) Clade B.

  20. In service inspection of fatigue cracks in a 1/R th scale PWR vessel with pressure cycles

    International Nuclear Information System (INIS)

    THE JOINT RESEARCH CENTER (JRC) at ISPRA has undertaken for several years a study on the propagation of fatigue cracks in PWR vessels. The goal is to establish a relation between the size and the location of the defects on the one hand, and the residual life of the structure on the other hand/1/. In order to verify the validity of theoretical models which were developed, a mock-up at a 1/5th scale of a PWR vessel was built. It was submitted to pressurization cycles representative of the operating conditions in a PWR. Therefore, it was necessary to have Non-Destructive methods which allows a stable and reliable track on a long period of time. This paper makes a review of results obtained by CEA during 5 successive In-Service Inspections on a 4 years period

  1. Validation of SWAT for burnup credit problems by analysis of post irradiation examination of 17*17 PWR fuel assembly

    International Nuclear Information System (INIS)

    For adopting burnup credit in transport or storage of spent fuel (SF), development of a reliable burnup calculation code is crucial. For this purpose, data of Post Irradiation Examination (PIE) have been extensively analyzed to evaluate accuracy of burnup calculation codes for a 14*14 or 15*15 PWR fuel assembly. This study shows results of analysis of this latest PIE with SWAT and ORIGEN2.1. SWAT is an integrated burnup code system for a 17*17 PWR fuel assembly that has been developed by Tohoku University and JAERI. The results show that SWAT can more precisely predict nuclide composition of latest PWR assembly than ORIGEN2.1. (O.M.)

  2. Valve inlet fluid conditions for pressurizer safety and relief valves in Westinghouse-designed plants. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Meliksetian, A.; Sklencar, A.M.

    1982-12-01

    The overpressure transients for Westinghouse-designed NSSSs are reviewed to determine the fluid conditions at the inlet to the PORV and safety valves. The transients considered are: licensing (FSAR) transients; extended operation of high pressure safety injection system; and cold overpressurization. The results of this review, presented in the form of tables and graphs, define the range of fluid conditions expected at the inlet to pressurized safety and power-operated relief valves utilized in Westinghouse-designed PWR units. These results will provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI/PWR Safety and Relief Valve Test Program indeed envelop those expected in their units.

  3. Air Abrasion

    Science.gov (United States)

    ... delivered directly to your desktop! more... What Is Air Abrasion? Article Chapters What Is Air Abrasion? What Happens? The Pros and Cons Will I Feel Anything? Is Air Abrasion for Everyone? print full article print this ...

  4. Guidelines - A Primer for Communicating Effectively with NABIR Stakeholders

    Energy Technology Data Exchange (ETDEWEB)

    A Harding; B Metting; C Word; G Bilyard; G Hund; J Amaya; J Weber; S Gajewski; S Underriner; T Peterson

    1998-12-10

    This primer is a tool to help prepare scientists for meetings with stakeholders. It was prepared for staff involved with the Natural and Accelerated Bioremediation Research (NABIR) program, sponsored by the U.S. Department of Energy. It discusses why some efforts in science communication may succeed while others fail, provides methods of approaching group interactions about science that may better orient expert participants, and summarizes experience drawn from observations of @oups interacting about topics in bioremediation or the NABIR program. The primer also provides briez usefid models for interacting with either expert or non-expert groups. Finally, it identifies topical areas that may help scientists prepare for public meetings, based on the developers' ongoing research in science communication in public forums.

  5. Guidelines - A Primer for Communicating Effectively with NABIR Stakeholders

    Energy Technology Data Exchange (ETDEWEB)

    Bilyard, G.R.; Word, C.J.; Weber, J.R.; Harding, A.K.

    2000-09-27

    This primer is a tool to help prepare scientists for meetings with stakeholders. It was prepared for staff involved with the Natural and Accelerated Bioremediation Research (NABIR) program, sponsored by the U.S. Department of Energy. It discusses why some efforts in science communication may succeed while others fail, provides methods of approaching group interactions about science that may better orient expert participants, and summarizes experience drawn from observations of groups interacting about topics in bioremediation or the NABIR program. The primer also provides brief, useful models for interacting with either expert or non-expert groups. Finally, it identifies topical areas that may help scientists prepare for public meetings, based on the developers' ongoing research in science communication in public forums.

  6. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  7. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Directory of Open Access Journals (Sweden)

    Benfarah Moez

    2016-01-01

    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

  8. Current status of research and development of nuclear fuel elements for PWR in Indonesia

    International Nuclear Information System (INIS)

    Full text: The energy need of Indonesia is increasing due to the population growth and for the economic progress. The government of Indonesia intends to apply an optimum energy mix comprising all viable prospective energy sources. The Government Regulation No. 5 year 2006 indicates the target of energy mix until 2025 and the share of nuclear energy is about 2% of primary energy or 4% of electricity (4000 MWe). The first two units of NPP is expected to be operated before 2020 as stated in Act No. 17 year 2007 on National Long Term Development Planning 2005-2025. The first NPP to be operated in Indonesia is PWR type with capacity of 1000 MWe/unit. One of the strategies to strength and increase national capacity in the program for NPP introduction is domestication industry for nuclear fuel. To reach this purpose, the activities of research and development is focus on nuclear fuel production technology for PWR. Currently research and development activity in Indonesia is to produce prototype of nuclear fuel element for PWR in the form of test fuel pin or mini pin. In this paper we will presenting the pelletization and fabrication technology development. The existing facility was designed for PHWR fuel element of CIRENE type. Development of pelletization technology is carried out by modifying the compacting machine. Parameters of compacting and sintering are determined based on both compressibility and compactibility of the pellet as indicated by density and mechanical strength of the UO2 green pellets. The sintering parameters to be determined are temperature, heating rate, and soaking time. Currently, the fabrication process is under experiment. All of the data resulted from the experiment that will be presented in this meeting. (author)

  9. Performance of PWR study in the technology supplier countries: south korea and japan case

    International Nuclear Information System (INIS)

    Electricity is needed as an infrastructure to support the national economic growth. For economic development sustainability, energy alternatives should be provided. Nuclear Power Plant (NPP) become the alternative of electricity generation for optimum energy mix in Indonesia and planned to operate in the 2016. Several studies have already done to prepare the NPP construction. This study focused on NPP performance especially PWR type in Asia, namely Japan and South Korea. Methodology used in this is literature tracing and a small calculation. The energy availability per unit per year is used as a parameter for evaluating the NPP performance. This conclusion are 1) the amount of NPP - PWR type in Japan is 22 units with total operational experiences 526 reactor-years and the average energy availability factor about 70.7% per unit per year. Meanwhile for the same type South Korea has 16 unit with total operational experience 222 reactor-years and average availability factor per unit per year is about 86.9%. 2) the 1000 class of PWR type both South Korea and Japan have 14 units. The operational experiences for thi class is 170 reactor-year for South Korean and 307 reactor-year for Japan. Meanwhile the average availability factor per unit per year is about 87.0% for South Korea and 69.6% for Japan. 3) the average availability factor is closed to capacity factor, so is important for real figure in assuming the techno-economic parameters, because it will influence the result o economic calculation. (author)

  10. PWR physics, operation and safety - Management of accidental situations of the reactor system

    International Nuclear Information System (INIS)

    This document contains a brief presentation and the table of contents of a book in which the author first presents the main types of accidents which are taken into account in safety demonstration. He presents the risk concerning the three safety barriers, and the various accidents affecting the three safety functions: reactivity control, power evacuation, confinement by the third barrier. Then the author describes approaches to the management of accidents affecting these three safety functions: reactivity insertion accidents due to absorber withdrawal (presentation, absorber cluster extraction transients, primary fluid dilution transient), steam pipe failure accidents or reactivity insertion by primary cooling (presentation, description of a transient of steam-pipe failure, sensitivity study of main parameters), loss-of-coolant accidents (presentation, intermediate breach, the big breach, peculiar case of breaches in stopped status), total loss of support systems such as in Fukushima (loss of electric supplies, of the cold source), loss of steam generator tubes. In the next part, the author addresses the Three Mile Island (TMI) accident and the lessons learned in terms of post-accidental management: presentation of the reactor and description of the accident. The author presents the 'status approach' of the post-accidental management, addresses the core post-fusion situations and their consequences as far as containment is concerned. He finally proposes ways to manage accidental situations for the PWR system. Appendices propose some additional aspects of system thermal-hydraulics, a presentation of safety deterministic and probabilistic approaches, comments on the Chernobyl and Fukushima accidents, comments on human and organizational factors regarding nuclear safety, some specific design aspects of the PWR reactor regarding safety, a presentation of assessment equations and data for the 1300 MWe PWR model

  11. Criticality analysis of PWR spent fuel storage facilities inside nuclear power plants

    International Nuclear Information System (INIS)

    This paper describes some of the main features of the actinide plus fission product burnup credit methodology used by Siemens for criticality safety design analysis of wet PWR storage pools with soluble boron in the pool water. Application of burnup credit requires knowledge of the isotopic inventory of the irradiated fuel for which burnup credit is taken. This knowledge is gained by using depletion codes. The results of the depletion analysis are a necessary input to the criticality analysis. Siemens performs depletion calculations for PWR fuel burnup credit applications with the aid of the Siemens standard design procedure SAV90. The quality of this procedure relies on statistics on the differences between calculation and measurement extracted from in-core measurement data and chemical assay data. Siemens performs criticality safety calculations with the aid of the criticality calculation modules of the SCALE code package. These modules are verified many times with the aid of various kinds of critical experiments and configurations: Application of these modules to spent LWR fuel assembly storage pools was verified by analyzing critical experiments simulating such storage pools. Actinide plus fission product burnup credit applications of these modules were verified by analyzing PWR reactor critical configurations. The result of performing a burnup credit analysis is the determination of a burnup, credit loading curve for the spent fuel storage racks designed for burnup credit. This curve specifies the loading criterion by indicating the minimum burnup necessary for the fuel assembly with a specific initial enrichment to be placed in the storage racks designed for burnup credit. The loading of the spent fuel storage racks designed for burnup credit requires the implementation of controls to ensure that the loading curve is met. The controls include the determination of fuel assembly burnup based on reactor records. (author)

  12. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    International Nuclear Information System (INIS)

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO2 fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  13. Cold leg condensation model for analyzing loss-of-coolant accident in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Jun, E-mail: liaoj@westinghouse.com; Frepoli, Cesare; Ohkawa, Katsuhiro

    2015-04-15

    Highlights: • Direct contact cold leg condensation model for full spectrum LOCA evaluation model. • The cold leg condensation model addresses both large break LOCA and small break LOCA. • The model is assessed against both large break and small break LOCA experiments. • Scalability of the cold leg condensation model to full scale PWR is discussed. - Abstract: Direct contact condensation in the cold leg of pressurized water reactor is an important phenomenon during a postulated loss-of-coolant accident. The amount of condensation in the cold legs impacts the thermal hydraulic behavior of the reactor coolant system and eventually the integration of reactor nuclear core. A cold leg condensation model was developed for the WCOBRA/TRAC-TF2 safety analysis code. The model correlated the COSI test data and addressed the scaling issues with respect to geometry, pressure, and steam and water flow rates expected during a typical PWR LOCA. The correlation was found to be in good agreement with separate effects and integral effects experimental data and implemented in the WCOBRA/TRAC-TF2 safety analysis code. The cold leg condensation model was assessed against various small break and large break LOCA separate effects tests such as COSI experiments, ROSA experiments and UPTF experiments. Those experiments cover a wide range of cold leg dimensions, system pressures, mass flow rates, and fluid properties. All the predicted condensation results match reasonably well with the experimental data. Scalability discussions on the diameter, flow area, length, superficial velocity, Reynolds number of both cold leg and SI line, and Froude number of SI line in the Westinghouse COSI test facility were provided. The distortion of the SI jet Reynolds number is moderate. The scaling analysis together with the validation matrix covering a wide range of cold leg diameter, SI flow rate and SI Reynolds number support the scalability of the developed cold leg condensation model to the full

  14. In- and ex-vessel coupled analysis of IVR-ERVC phenomenon for large scale PWR

    International Nuclear Information System (INIS)

    Highlights: • MELCOR models are built for large scale PWR with thermal power reaching 5000 MWt. • In- and ex-vessel coupled transient analysis of IVR-ERVC phenomenon is performed. • Results show that the IVR-ERVC strategy is an effective way to maintain RPV integrity during a severe accident. - Abstract: As a key severe accident management strategy for light water reactors (LWRs), in-vessel retention (IVR) through external reactor vessel cooling (ERVC) has been the focus of relevant studies for decades. However, previous studies only investigated the molten pool configurations considered to be in a final steady state mainly for reactors of such as AP600 and AP1000. Furthermore, most of studies performed in the past dealt with analysis for an isolated IVR-ERVC process, without considering the strong coupling between the internal and external reactor pressure vessel (RPV) conditions. This paper addresses the IVR-ERVC issues from a transient perspective using the severe accident code MELCOR for a large advanced passive power plant: a three-loop, 5000 MWt scale pressurized water reactor with passive safety features. The analysis is mainly focused on the severe accident transients including core degradation and relocation, molten pool formation and growth, and heat transfer within a molten pool. Furthermore, internal and external RPV conditions are combined together in the IVR-ERVC analysis. MELCOR calculations for lower head heat flux are then compared with critical heat flux (CHF) to assess the effectiveness of IVR-ERVC. The results suggest that lower head heat flux is below the CHF value. Therefore, the IVR-ERVC strategy for this large PWR is considered to be feasible. It was also found that as the reactor power is raised to large scale PWR, new accident sequences may occur during the severe accident evolution, thus leading to a proposal of a completely new molten pool configuration for future studies

  15. PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Blakeman, Edward D [ORNL; Peplow, Douglas E. [ORNL; Wagner, John C [ORNL; Murphy, Brian D [ORNL; Mueller, Don [ORNL

    2007-09-01

    The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.

  16. Introduction to Command, Control, and Communications: a primer

    OpenAIRE

    Eidson, Edward H.

    1995-01-01

    This thesis is a primer for students in the Introduction to Command, Control, and Communications Courses at the Naval Postgraduate School. This document provide students a consolidated reference that emphasizes key concepts and ideas presented by the course instructor and required readings. Its organization closely parallels the course outline used for in-class instruction. It supports course objectives by providing an executive overview of a wide variety of C2 topics. These topics include: C...

  17. Fundamentals of statistical evidence—a primer for legal professionals

    OpenAIRE

    Aitken, Colin; TARONI, FRANCO

    2008-01-01

    Criminal courts are today frequently confronted with statistical evidence, notably in relation to DNA profiling. Recent experience tends to confirm both widespread perceptions and more systematic research indicating that probability and statistics are not handled confidently, or always competently, by lawyers, judges, jurors or even by forensic scientists. Conceived as a primer for legal professionals, this article reviews basic statistical terminology and its forensic applications, and explo...

  18. Small Commercial Building Re-tuning: A Primer

    Energy Technology Data Exchange (ETDEWEB)

    Cort, Katherine A.; Hostick, Donna J.; Underhill, Ronald M.; Fernandez, Nicholas; Katipamula, Srinivas

    2013-09-30

    To help building owners and managers address issues related to energy-efficient operation of small buildings, DOE has developed a Small Building Re-tuning training curriculum. This "primer" provides additional background information to understand some of the concepts presented in the Small Building Re-tuning training. The intent is that those who are less familiar with the buidling energy concepts will review this material before taking the building re-tuning training class.

  19. Primer on Molecular Genetics; DOE Human Genome Program

    Science.gov (United States)

    1992-04-01

    This report is taken from the April 1992 draft of the DOE Human Genome 1991--1992 Program Report, which is expected to be published in May 1992. The primer is intended to be an introduction to basic principles of molecular genetics pertaining to the genome project. The material contained herein is not final and may be incomplete. Techniques of genetic mapping and DNA sequencing are described.

  20. MERS-CoV PCR/Sequencing Primers

    OpenAIRE

    sprotocols

    2014-01-01

    Authors: Rachel Graham ### Abstract This protocol details the primers and conditions used for forward and reverse PCR amplification and sequencing of MERS-CoV genomes. ### Introduction This protocol details the steps, reagents, and conditions required to sequence MERS-CoV genomes in the forward and reverse directions. The protocol begins with the RT-PCR step, assuming that MERS-CoV RNA purified using a standard procedure (i.e., TRIzol extraction) has already been performed. ...

  1. The Primer for Sports Medicine Professionals on Imaging

    OpenAIRE

    Farshad-Amacker, Nadja A.; Jain Palrecha, Sapna; Farshad, Mazda

    2013-01-01

    Because of its inherent superior soft tissue contrast and lack of ionizing radiation, magnetic resonance imaging (MRI) is highly suited to study the complex anatomy of the shoulder joint, particularly when assessing the relatively high incidence of shoulder injuries in young, athletic patients. This review aims to serve as a primer for understanding shoulder MRI in an algorithmical approach, including MRI protocol and technique, normal anatomy and anatomical variations of the shoulder, pathol...

  2. Primer on molecular genetics. DOE Human Genome Program

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This report is taken from the April 1992 draft of the DOE Human Genome 1991--1992 Program Report, which is expected to be published in May 1992. The primer is intended to be an introduction to basic principles of molecular genetics pertaining to the genome project. The material contained herein is not final and may be incomplete. Techniques of genetic mapping and DNA sequencing are described.

  3. Perbedaan Kadar Superokside Dismutase pada Remaja dengan Dismenore Primer dan Tanpa Dismenore Primer

    Directory of Open Access Journals (Sweden)

    Yanti .

    2016-01-01

    Full Text Available Abstrak             Dismenore didefinisikan sebagai rasa kram saat menstruasi yang menyakitkan tanpa patologi yang jelas. Kram berlangsung selama satu hari atau lebih dan disertai rasa mual, diare, sakit kepala. Masalah yang ditimbulkan oleh dismenore adalah  peningkatan ketidakhadiran di sekolah pada remaja sehingga menyebabkan rendahnya nilai akademik pada pelajar. Superokside dismutase (SOD adalah bahan bioaktif yang diketahui bersifat antioksidan. SOD melindungi sel terhadap gangguan oksidan (radikal bebas. SOD mengubah anion superoksida menjadi hidrogen peroksida dan oksigen, sering disebut juga sebagai pertahanan primer terhadap stress oksidatif. Tujuan penelitian ini adalah mengetahui  perbedaan kadar superokside dismutase pada remaja dengan dismenore dan tanpa dismenore. Penelitian ini adalah observasional desain cross sectional comparative. Data dianalisis menggunakan uji Mann-Withney  dengan nilai p<0.05 dianggap bermakna secara statistik. Rerata kadar SOD pada remaja yang mengalami dismenore yaitu 36,76 u/ml dan rerata kadar SOD pada remaja tanpa dismenore yaitu 32,24 u/ml. Dengan nilai p>0,005 (0,345. Hasil penelitian ini menyimpulkan bahwa tidak terdapat perbedaan yang bermakna  kadar SOD pada remaja dengan dismenore dan tanpa dismenore. Kata kunci: remaja, dismenore, antioksidan, superokside dismutase AbstractPrimary dysmenorrhoe is  a painful menstrual cramps without obvious pathology. Cramps is lasting for one day or more, accaompanied by nausea, diarrhea and headache. Problems cause by dysmenorrhea are an increase in school attendance in adolescents resulting in low academic grades of students. Superokside Dismeutase (SOD is a bioactive ingredient that is known as antioxidants, protecting cells against harmful SOD oxidants (free radicals SOD convert superoxide anion into hydrogen perokxide and oxygen, often call  as primary defense agains oxidative stress. Primary dysmenorrhoe increased uterine activity or

  4. Sensory reception of the primer pheromone ethyl oleate

    Science.gov (United States)

    Muenz, Thomas S.; Maisonnasse, Alban; Plettner, Erika; Le Conte, Yves; Rössler, Wolfgang

    2012-05-01

    Social work force distribution in honeybee colonies critically depends on subtle adjustments of an age-related polyethism. Pheromones play a crucial role in adjusting physiological and behavioral maturation of nurse bees to foragers. In addition to primer effects of brood pheromone and queen mandibular pheromone—both were shown to influence onset of foraging—direct worker-worker interactions influence adult behavioral maturation. These interactions were narrowed down to the primer pheromone ethyl oleate, which is present at high concentrations in foragers, almost absent in young bees and was shown to delay the onset of foraging. Based on chemical analyses, physiological recordings from the antenna (electroantennograms) and the antennal lobe (calcium imaging), and behavioral assays (associative conditioning of the proboscis extension response), we present evidence that ethyl oleate is most abundant on the cuticle, received by olfactory receptors on the antenna, processed in glomeruli of the antennal lobe, and learned in olfactory centers of the brain. The results are highly suggestive that the primer pheromone ethyl oleate is transmitted and perceived between individuals via olfaction at close range.

  5. Teaching Thermal Hydraulics & Numerical Methods: An Introductory Control Volume Primer

    Energy Technology Data Exchange (ETDEWEB)

    D. S. Lucas

    2004-10-01

    A graduate level course for Thermal Hydraulics (T/H) was taught through Idaho State University in the spring of 2004. A numerical approach was taken for the content of this course since the students were employed at the Idaho National Laboratory and had been users of T/H codes. The majority of the students had expressed an interest in learning about the Courant Limit, mass error, semi-implicit and implicit numerical integration schemes in the context of a computer code. Since no introductory text was found the author developed notes taught from his own research and courses taught for Westinghouse on the subject. The course started with a primer on control volume methods and the construction of a Homogeneous Equilibrium Model (HEM) (T/H) code. The primer was valuable for giving the students the basics behind such codes and their evolution to more complex codes for Thermal Hydraulics and Computational Fluid Dynamics (CFD). The course covered additional material including the Finite Element Method and non-equilibrium (T/H). The control volume primer and the construction of a three-equation (mass, momentum and energy) HEM code are the subject of this paper . The Fortran version of the code covered in this paper is elementary compared to its descendants. The steam tables used are less accurate than the available commercial version written in C Coupled to a Graphical User Interface (GUI). The Fortran version and input files can be downloaded at www.microfusionlab.com.

  6. Effect of oligonucleotide primers in determining viral variability within hosts

    Directory of Open Access Journals (Sweden)

    Moya Andrés

    2004-12-01

    Full Text Available Abstract Background Genetic variability in viral populations is usually estimated by means of polymerase chain reaction (PCR based methods in which the relative abundance of each amplicon is assumed to be proportional to the frequency of the corresponding template in the initial sample. Although bias in template-to-product ratios has been described before, its relevance in describing viral genetic variability at the intrapatient level has not been fully assessed yet. Results To investigate the role of oligonucleotide design in estimating viral variability within hosts, genetic diversity in hepatitis C virus (HCV populations from eight infected patients was characterised by two parallel PCR amplifications performed with two slightly different sets of primers, followed by cloning and sequencing (mean = 89 cloned sequences per patient. Population genetics analyses of viral populations recovered by pairs of amplifications revealed that in seven patients statistically significant differences were detected between populations sampled with different set of primers. Conclusions Genetic variability analyses demonstrates that PCR selection due to the choice of primers, differing in their degeneracy degree at some nucleotide positions, can eclipse totally or partially viral variants, hence yielding significant different estimates of viral variability within a single patient and therefore eventually producing quite different qualitative and quantitative descriptions of viral populations within each host.

  7. The continued development of the MFM suite and its practical application on a PWR system

    DEFF Research Database (Denmark)

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies in phys...... physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

  8. Recent bibliography on analytical and sampling problems of a PWR primary coolant Suppl. 3

    International Nuclear Information System (INIS)

    The present supplement to the bibliography on analytical and sampling problems of PWR primary coolant covers the literature published in 1984 and includes some references overlooked in the previous volumes dealing with the publications of the last 10 years. References are devided into topics characterized by the following headlines: boric acid; chloride; chlorine; carbon dioxide; general; gas analysis; hydrogen isotopes; iodine; iodide; nitrogen; noble gases and radium; ammonia; ammonium; oxygen; other elements; radiation monitoring; reactor safety; sampling; water chemistry. Under a given subject bibliographical information is listed in alphabetical order of the authors. (V.N.)

  9. The three-dimensional PWR transient code ANTI; rod ejection test calculation

    International Nuclear Information System (INIS)

    ANTI is a computer program being developed for three-dimensional coupled neutronics and thermal-hydraulics description of a PWR core under transient conditions. In this report a test example calculated by the program is described. The test example is a simulation of a control rod ejection from a very small reactor core (to save somputing time). In order to show the influence of cross flow between adjacent fuel elements the same calculation was performed both with the cross flow option and with closed hydraulic channels. (author)

  10. Simulation of a low-pressure severe accident scenario in a PWR with ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Mathias; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2013-07-01

    The plant behavior of a Pressurized Water Reactor (PWR) during a severe accident scenario is analyzed with system code ATHLET-CD Mod. 2.2C in order to assess the code capabilities in terms of the late-phase of the core degradation. For this purpose a severe accident sequence caused by a Station Black-out and a large break in the primary cooling system is simulated both without any accident management measures and with a delayed reflooding of the substantially degraded core. Selected code results are presented in this paper. (orig.)

  11. Application of LBB to high energy piping systems in operating PWR

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  12. Simulation model for the dynamic behavior of the hydraUlic circuito of PWR reactors

    International Nuclear Information System (INIS)

    The present work consist of the development of a computer code for the simulations of hydraulic transients caused by stoppages of the primary coolant pumps of nuclear reactors and it applied to the hydraulic circuits typical of PWR reactor. The code calculates the time-histories of the mass flux, rotation speed, electric and hydraulic torque and dynamic head of the pumps. It can be used for any combination of active and inactive pumps. Several transients were analysed and the results were compared with comparared with data from the Angra-I nuclear power plant. The results were considered satisfactory. (author)

  13. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  14. Environment-insensitive equivalent diffusion theory group constants for PWR [pressurized water reactor] radial reflector regions

    International Nuclear Information System (INIS)

    A novel approach combining the nodal equivalence theory (NET) and the response matrix homogenization methods into a single procedure, has been developed for generating equivalent nodal parameters for the radial reflector of a PWR [pressurized water reactor]. This procedure yields equivalent parameters which are per definition less sensitive to local environment than those obtained via a straightforward NET homogenization method. Unlike other response matrix approaches which may achieve this same goal, this method requires only a single multi-group spectral calculation to obtain all the necessary information for generating the equivalent parameters. Numerical results are presented which demonstrate the potential and the superiority of this reflector model over other modern models

  15. A systematic approach for development of a PWR cladding corrosion model

    International Nuclear Information System (INIS)

    A new model for the in-reactor corrosion of Improved (low-tin) Zircaloy-4 cladding irradiated in commercial pressurized water reactors (PWRs) is described. The model is based on an extensive database of PWR fuel cladding corrosion data from fuel irradiated in commercial reactors, with a range of fuel duty and coolant chemistry control strategies which bracket current PWR fuel management practices. The fuel thermal duty with these current fuel management practices is characterized by a significant amount of sub-cooled nucleate boiling (SNB) during the fuel's residence in-core, and the cladding corrosion model is very sensitive to the coolant heat transfer models used to calculate the coolant temperature at the oxide surface. The systematic approach to developing the new corrosion model therefore began with a review and evaluation of several alternative models for the forced convection and SNB coolant heat transfer. The heat transfer literature is not sufficient to determine which of these heat transfer models is most appropriate for PWR fuel rod operating conditions, and the selection of the coolant heat transfer model used in the new cladding corrosion model has been coupled with a statistical analysis of the in-reactor corrosion enhancement factors and their impact on obtaining the best fit to the cladding corrosion data. The in-reactor corrosion enhancement factors considered in this statistical analysis are based on a review of the current literature for PWR cladding corrosion phenomenology and models. Fuel operating condition factors which this literature review indicated could have a significant effect on the cladding corrosion performance were also evaluated in detail in developing the corrosion model. An iterative least squares fitting procedure was used to obtain the model coefficients and select the coolant heat transfer models and in-reactor corrosion enhancement factors. This statistical procedure was completed with an exhaustive analysis of the model

  16. The continued development of the MFM suite and its practical application on a PWR system

    DEFF Research Database (Denmark)

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies...... in physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

  17. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  18. Application of fine absolute rated filters to PWR CVCS and radwaste filter duties

    International Nuclear Information System (INIS)

    Filters for PWR systems have generally been specified in terms of nominal removal ratings. Products supplied by different manufacturers against these specifications vary considerably in removal efficiency and service life. The paper describes the method used by Pall Corporation for reliable, reproducible assessment of filter performance and determination of absolute filtration ratings. The old nominal specifications are shown to be too coarse for effective removal of particle borne activity. The introduction into service of new absolute rated fine filter media with positive zeta potential is described. (author)

  19. LOBI-BL-40: an approach to the SGTR phenomenology in a one loop PWR

    International Nuclear Information System (INIS)

    The aim of the ''LOBI-BL-40'' project is to study the phenomena and the recovery techniques involved in Steam Generator Tube Rupture (SGTR) events in a commercial single-loop PWR plant. Experimental support for the project will be provided by the BL-40 experiment, to be carried out at the LOBI installation (JRC, Ispra-Varese, Italy). This paper describes the project, its objectives, the activities involved and the progress made, with special attention to specification of the BL-40 experiments. The work is being undertaken by the Spanish utility UNION-FENOSA. (author)

  20. Application of burnup credit with partial boron credit to PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    The outcome of performing a burnup credit criticality safety analysis of a PWR spent fuel storage pool is the determination of burnup credit loading curves BLC=BLC(e) for the spent fuel storage racks designed for burnup credit, cp. Reference. A burnup credit loading curve BLC=BLC(e) specifies the loading criterion by indicating the minimum burnup BLC(e) necessary for the fuel assembly with a specific initial enrichment e to be placed in storage racks designed for burnup credit. (orig.)

  1. Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT

    Directory of Open Access Journals (Sweden)

    Uwe Imke

    2012-01-01

    Full Text Available SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT. The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

  2. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  3. Guangdong: China imports French and British technology for a twin unit PWR station

    International Nuclear Information System (INIS)

    A series of articles on the Guangdong twin unit PWR station at Daya Bay in China. The Guangdong Nuclear Power Joint Venture Company (GNPJVC) is the joint Chinese-Hong Kong grouping responsible for owning and operating the Guangdong station. Contracts have been signed with Framatome, EdF and GEC Turbine Generators for the supply of two 900MWe PWRs with associated turbine generators and ancillary equipment, and relevant training and technology transfer. The meeting of site specific needs is discussed and a description of the plant given. A cutaway diagram together with technical specifications of the Guangdong plant is presented. (U.K.)

  4. Behaviour of a PWR safety containment with rising internal pressure load

    International Nuclear Information System (INIS)

    For the PWR Philippsburg (KKP II) a slowly increasing containment pressure typical for a core melt accident is assumed. The failure pressure and failure mode of the containment are determined using the ROTMEM code. It turns out that the sealing box covering the bolted connection at the equipment hatch will fail at a containment overpressure between 12.9 and 13.7 bar. Then the leakage through the bolted connection is sufficient to prevent further pressure increase. However, if the sealing box failed at a somewhat higher pressure, a global containment failure with extreme mechanical damage would have to be expected. (orig.)

  5. Logic flowgraph model for disturbance analysis of a PWR pressurizer system

    Energy Technology Data Exchange (ETDEWEB)

    Guarro, S.; Okrent, D.

    1984-01-01

    The Logic Flowgraph Methodology (LFM) has been developed as a synthetic simulation language for process reliability or disturbance analysis applications. A Disturbance Analysis System (DAS) using the LFM models can store the necessary information concerning a given process in an efficient way, and automatically construct in real time the diagnostic tree(s) showing the root cause(s) of occurring disturbances. A comprehensive LFM model for a PWR pressurizer system is presented and discussed, and the latest version of the LFM tree synthesis routine, optimized to achieve reduction of computer memory usage, is used to show the LFM diagnoses of selected hypothetic disturbances.

  6. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  7. Validation of seismic soil structure interaction (SSI) methodology for a UK PWR nuclear power station

    International Nuclear Information System (INIS)

    The seismic loading information for use in the seismic design of equipment and minor structures within a nuclear power plant is determined from a dynamic response analysis of the building in which they are located. This dynamic response analysis needs to capture the global response of both the building structure and adjacent soil and is commonly referred to as a soil structure interaction (SSI) analysis. NNC have developed a simple and cost effective methodology for the seismic SSI analysis of buildings in a PWR nuclear power station at a UK soft site. This paper outlines the NNC methodology and describes the approach adopted for its validation

  8. The effects of fission gas release on PWR fuel rod design and performance

    International Nuclear Information System (INIS)

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50 000 to 60 000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  9. The effects of fission gas release on PWR fuel rod design and performance

    International Nuclear Information System (INIS)

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50,000 to 60,000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  10. The analysis of dose reduction practices in the PWR units of Electricite de France

    International Nuclear Information System (INIS)

    Electricite de France is currently operating twenty-one 900 MWe standard PWR units. Along with the increasing size of the french nuclear program, experience has been gathered on the problem of occupational exposures and actions have been undertaken in order to keep doses of workers as low as reasonably achievable. This paper first presents the main results concerning collective exposures at units already in operation. Then a set of protective actions already taken have been assessed, taking into account their costs and effectiveness and allowing for the possible interdependence of protection and energy production objectives. The limits of such an approach are discussed in conclusion. (author)

  11. Evolution of reactor monitoring and protection systems for PWR; Evolution des systemes de surveillance et de protection des REP

    Energy Technology Data Exchange (ETDEWEB)

    Chaloin, B. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Mourlevat, J.L. [FRAMATOME ANP, 92 - Paris-La-Defence (France)

    2004-07-01

    This paper presents the evolution of the reactor protection systems and of the reactor monitoring systems for PWR since the initial design in the Fessenheim plant to the latest development for the EPR (European pressurized reactor). The features of both systems for the different kinds of PWR operating in France: 900 MWe, 1300 MWe and N4, are reviewed. The expected development of powerful micro-processors for computation, for data analysis and data storage will make possible in a near future the monitoring on a 3-dimensional basis and on a continuous manner, of the nuclear power released in the core. (A.C.)

  12. Assessment of management alternatives for LWR wastes. Volume 3. Description of German scenarios for PWR and BWR wastes

    International Nuclear Information System (INIS)

    This report deals with the description of a management route for PWR waste relying to a certain extent on German practices in this particular area. This description, which aims at providing input data for subsequent cost evaluation, includes all management steps which are usually implemented for solid, liquid and gaseous wastes from their production up to the interim storage of the final waste products. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for PWR and BWR wastes based on economical and radiological criteria

  13. Recovery of the mitochondrial COI barcode region in diverse Hexapoda through tRNA-based primers

    Directory of Open Access Journals (Sweden)

    Oh Hyun-Woo

    2010-07-01

    Full Text Available Abstract Background DNA barcoding uses a 650 bp segment of the mitochondrial cytochrome c oxidase I (COI gene as the basis for an identification system for members of the animal kingdom and some other groups of eukaryotes. PCR amplification of the barcode region is a key step in the analytical chain, but it sometimes fails because of a lack of homology between the standard primer sets and target DNA. Results Two forward PCR primers were developed following analysis of all known arthropod mitochondrial genome arrangements and sequence alignment of the tRNA-W gene which was usually located within 200 bp upstream of the COI gene. These two primers were combined with a standard reverse primer (LepR1 to produce a cocktail which generated a barcode amplicon from 125 of 141 species that included representatives of 121 different families of Hexapoda. High quality sequences were recovered from 79% of the species including groups, such as scale insects, that invariably fail to amplify with standard primers. Conclusions A cocktail of two tRNA-W forward primers coupled with a standard reverse primer amplifies COI for most hexapods, allowing characterization of the standard barcode primer binding region in COI 5' as well as the barcode segment. The current results show that primers designed to bind to highly conserved gene regions upstream of COI will aid the amplification of this gene region in species where standard primers fail and provide valuable information to design a primer for problem groups.

  14. Sequence-related amplified polymorphism primer screening on Chinese fir (Cunninghamia lanceolata (Lamb.) Hook)

    Institute of Scientific and Technical Information of China (English)

    Huiquan Zheng; Hongjing Duan; Dehuo Hu; Ruping Wei; Yun Li

    2015-01-01

    Chinese fir (Cunninghamia lanceolata (Lamb.) Hook) is one of the most important coniferous tree species used for timber production in China. Here, we conducted a sequence-related amplified polymorphism (SRAP) primer screening assay with a total of 594 primer combinations, using 22 forward and 27 reverse primers on four repre-sentative Chinese fir genotypes. The obtained results indicated that Chinese fir genomic DNA has a notable amplification bias on the employed forward or reverse primer nucleotides (3' selection bases). Out of the tested primer sets, 35 primer combinations with clearly distin-guished bands, stable amplification, and rich polymorphism were selected and identified as optimal primer sets. These optimal primer pairs gave a total of 379 scorable bands, including 265 polymorphic bands, with an average of 10.8 bands and 7.6 polymorphic bands per primer combination. The produced band number for each optimal primer set ranged from 7 to 14 with a percentage of polymorphic bands spanning from 33.3 to 100.0%. These primer combinations could facilitate the next SRAP analysis assays in Chinese fir.

  15. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  16. Estimate of the speed of the refrigerant on a PWR: three way based on the analysis of noise; Estimacion de la volecidad del refrigerante en un PWR: tres vias basadas en el analisis de ruido

    Energy Technology Data Exchange (ETDEWEB)

    Montalvo, C.; Ruiz, M.; Garcia Berrocal, A.

    2014-07-01

    The speed of the refrigerant is a key parameter in the monitoring of the operation a PWR. He know this value and be able to track on-site It allows an understanding of the State of the kernel with valuable information about the refrigerant, and thus behavior on heat exchange which takes place in the reactor. (Author)

  17. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    Institute of Scientific and Technical Information of China (English)

    XU Mi

    2007-01-01

    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  18. Computer code validation study of PWR core design system, CASMO-3/MASTER-{alpha}

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Kim, M. H. [Kyounghee Univ., Taejon (Korea, Republic of); Woo, S. W. [KINS, Taejon (Korea, Republic of)

    1999-05-01

    In this paper, the feasibility of CASMO-3/MASTER-{alpha} nuclear design system was investigated for commercial PWR core. Validation calculation was performed as follows. Firstly, the accuracy of cross section generation from table set using linear feedback model was estimated. Secondly, the results of CASMO-3/MASTER-{alpha} was compared with CASMO-3/NESTLE 5.02 for a few benchmark problems. Microscopic cross sections computed from table set were almost the same with those from CASMO-3. There were small differences between calculated results of two code systems. Thirdly, the repetition of CASMO-3/MASTER-{alpha} calculation for Younggwang Unit-3, Cycle-1 core was done and their results were compared with nuclear design report(NDR) and uncertainty analysis results of KAERI. It was found that uncertainty analysis results were reliable enough because results were agreed each other. It was concluded that the use of nuclear design system CASMO-3/MASTER-{alpha} was validated for commercial PWR core.

  19. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  20. Phenomenology and Course of Severe Accidents in PWR-Plants - Training by Teaching and Demonstration

    International Nuclear Information System (INIS)

    A special one day training course on 'Phenomenology and Course of Severe Accidents in PWR-Plants' was developed at GRS initiated by the interest of German utilities. The work was done in the frame of projects sponsored by the German Ministries for Environment, Nature Conservation and Nuclear Safety (BMU) and for Education, Science, Research and Technology (BMBF). In the paper the intention and the subject of this training course will be discussed and selected parts of the training course will be presented. Demonstrations are made within this training course with the GRS simulator system ATLAS to achieve a broader understanding of the phenomena discussed and the propagation of severe accidents on a plant specific basis. The GRS simulator system ATLAS is linked in this case to the integral code MELCOR and pre-calculated plant specific severe accident calculations are used for the demonstration together with special graphics showing plant specific details. Several training courses have been held since the first one in November, 1996 especially to operators, shift personal and the management board of a German PWR. In the meantime the training course was updated and suggestions for improvements from the participants were included. In the future this type of the training course will be made available for members of crisis teams, instructors of commercial training centres and researchers of different institutions too. (authors)

  1. The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 1019n.cm-2 (E>1.0MeV) at 250-2900C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu3P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

  2. Optimal burnable poison-loading in a PWR with carbon coated particle fuel

    International Nuclear Information System (INIS)

    An innovative PWR concept that uses carbon-coated particle fuels moderated by graphite as that of HTGR but cooled by pressurized light water has been studied. The aim of this concept is to take both the best advantages of fuel integrity against fission products release and the reliability PWR technology based on the long operational experience. The purpose of the study is to optimize loading pattern of burnable poison in the proposed core in order to suppress excess reactivity during a cycle. Although there are many parameters to be determined for optimization of the usage of burnable poison, the emphasis is put here on loading patterns of Gadolinia in an assembly and in the core. We investigated the burnup characteristics of the core varying the concentration of burnable poison in a fuel rod, the number of burnable poison-rods in an assembly, and the number of burnable poison-assemblies in the core. The result suggested that Gadolinia was more suitable for this reactor than boron as burnable poison, and it was possible to make the reactivity swing negligible by combining at least three kinds of burnable poison-assemblies in which the amount of Gadolinia was different. Therefore the requirement for the number of control rods was reduced and it meant that Control Rod Programming would become easier. (author)

  3. Validation of STAR-CCM+ for bouyancy driven mixing in a PWR reactor pressure vessel

    International Nuclear Information System (INIS)

    Within the OECD/NEA PKL-II project, experiments have been carried out aimed at investigating the flow mixing in the downcomer and lower plenum of a pressurized water reactor (PWR) in the buoyancy driven mixing regimes. The experiments have been performed at the ROCOM test facility, a 1:5 scaled representation of a KONVOI type pressurized water reactor (PWR). The facility is equipped with advanced instrumentation (i.e. wiremesh sensors) allowing a detailed measurement of flow mixing in the downcomer annulus and at the core inlet. A computational fluid dynamic (CFD) model has been developed at the Paul Scherrer Institute within the STARS project, employing the STAR-CCM+ code. The CFD model has been validated against the ROCOM experimental results. It has been shown that the developed model provided a good agreement with experiment. In order to evaluate the difference between momentum driven and density driven mixing regimes, calculations were performed assuming no density difference, and with 12% higher density in one of the loops respectively. (author)

  4. ALIBABA, an assistance system for the detection of confinement leaks in a PWR reactor

    International Nuclear Information System (INIS)

    The objective of the Crisis Technical Center (CTC) of the French Institute for Nuclear Protection and Safety (IPSN) is to estimates the consequences of a given nuclear accident on the populations and the environment. ALIBABA is a data processing tool available at the CTC and devoted to the detection of confinement leaks in 900 MWe PWR reactors using the activity values measured by the captors of the installation. The heart of this expert system is a structural and functional representation of the different components directly involved in the leak detection (isolating valves, ventilation systems, electric boards etc..). This tool can manage the availability of each component to make qualitative and quantitative balance-sheets. This paper presents the ALIBABA software, an industrial prototype realized with the SPIRAL knowledge base systems generator at the CEA Reactor Studies and Applied Mathematics Service (SERMA) and commercialized by CRIL-Ingenierie Society. It describes the techniques used for the modeling of PWR systems and for the visualization of the survey. The functionality of the man-machine interface is discussed and the means used for the validation of the software are summarized. (J.S.). 6 refs

  5. Design features of Tomari unit 3. Latest 900MW-class PWR design in Japan

    International Nuclear Information System (INIS)

    Tomari Unit 3 is the latest 912 MW PWR in Japan. Its major design features are (1) improvement of thermal efficiency, (2) application of the advanced main control board, (3) system design according to PSA, (4) fatigue analysis in consideration of environmental effect and high-cycle thermal fatigue and (5) concurrent engineering and 3D-CAD. The improvement of thermal efficiency is achieved by increasing generating capacity. It is the result of increase of reactor coolant temperature, improvement of steam turbine efficiency and higher vacuum of condenser. The improvement of steam turbine efficiency is the result of 3D hydraulic design and 54-inch last blade of LP turbine. 54-inch last blade is the first application in Japan. The advanced main control board is the first application to PWR in Japan and reduces operator's mental workload and human error probability. System design according to PSA, fatigue analysis in consideration of environmental effect and high-cycle thermal fatigue and 3D-CAD enhance safety, reliability and integrity of Tomari Unit 3. (author)

  6. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  7. Nuclear criticality safety analysis for the traveller PWR fuel shipping package

    Energy Technology Data Exchange (ETDEWEB)

    Vescovi, P.J.; Kent, N.A.; Casado, C.A. [Westinghouse Electric Co., LLC, Columbia, SC (United States)]|[ENUSA Industrias Avanzadas SA, Madrid (Spain)

    2004-07-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse.

  8. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  9. Application of liquid chromatography techniques to the measurement of soluble transition metals in PWR primary coolant

    International Nuclear Information System (INIS)

    Two chromatographic techniques have been developed, and evaluated for the on-line analysis of soluble transition metals, particularly cobalt, in PWR primary coolant. Automatic operation and control, together with data processing and storage has been achieved by interfacing a Dionex ion chromatograph to a microprocessor control system. An absolute detection limit of 0.1 ng cobalt has been obtained which, with on-line sample preconcentration (100 ml), has enabled measurements to be made down to part-per-trillion levels (0.001 ppb). Application of the techniques to PWR coolant analysis was demonstrated by a programme of work on the Half Megawatt Loop at Winfrith. During this work some aspects of the behaviour of soluble metal species have been studied in both de-oxygenated and hydrogenated conditions. The effects of changes in coolant chemistry, operating temperature, and sample line flowrates on circulating impurity levels are reported, together with the dramatic effects observed when part of the circuit pipework was replaced with new stainless steel tubing. (author)

  10. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  11. Nuclear criticality safety analysis for the traveller PWR fuel shipping package

    International Nuclear Information System (INIS)

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse

  12. Evaluation of new pH control agents for PWR secondary water systems. Final report

    International Nuclear Information System (INIS)

    This report presents an account of work evaluating new organic pH control agents for PWR station secondary systems. A discussion of corrosion control with pH controlling amines is given together with some information on current utility practice with volatile amines. The report describes opportunities for improved corrosion control with new pH controlling amines and the selection of candidates for laboratory evaluation. Progress on EPRI program RP 1571-7, ''Spectroscopic Characterization Methods for New pH Control Agents'', is presented. Also presented is an outline of the second phase of the RP 1571-7 program which was directed toward ''Evaluation of pH Control Agents in Simulated Steam Generator Tubesheet Crevice Environments.'' Quinuclidine and diaminopropane were evaluated using laser Raman spectroscopy. Hydrothermal stability and pH properties at temperature were determined. Quinuclidine was superior to diaminopropane and presents advantages over ammonia, hydrazine and morpholine for plant use as a pH control agent. It was further evaluated in tests with a scaled PWR crevice simulator for corrosion compatibility as well as pH control. 17 refs., 22 figs., 9 tabs

  13. Development and preliminary verification of the PWR on-line core monitoring software system. SOPHORA

    International Nuclear Information System (INIS)

    This paper presents an introduction to the development and preliminary verification of a new on-line core monitoring software system (CMSS), named SOPHORA, for fixed in-core detector (FID) system of PWR. Developed at China General Nuclear Power Corporation (CGN), SOPHORA integrates CGN’s advanced PWR core simulator COCO and thermal-hydraulic sub-channel code LINDEN to manage the real-time core calculation and analysis. Currents measured by the FID are re-evaluated and used as bases to reconstruct the 3-D core power distribution. The key parameters such as peak local power margin and minimum DNBR margin are obtained by comparing with operation limits. Pseudo FID signals generated by data from movable in-core detector (MID) are used to verify the SOPHORA system. Comparison between predicted power peak and the responding MID in-core flux map results shows that the SOPHORA results are reasonable and satisfying. Further verification and validation of SOPHORA is undergoing and will be reported later. (author)

  14. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Hermann, O.W.

    1996-09-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  15. Power-cooling mismatch test series. Test PCM-2A; test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cawood, G.W.; Holman, G.W.; Martinson, Z.R.; Legrand, B.L.

    1976-09-01

    The report describes the results of an in-pile experimental investigation of pre- and postcritical heat flux (CHF) behavior of a single 36-inch-long, pressurized water reactor (PWR) type, UO/sub 2/-fueled, zircaloy-clad fuel rod. The nominal coolant conditions for pressure and temperature were representative of those found in a commercial PWR. Nine separate departures from nucleate boiling (DNB) cycles were performed by either of two different methods: (a) decreasing the coolant flow rate while the fuel rod power was held constant, or (b) increasing the fuel rod power while the coolant flow rate was held constant. DNB was obtained during eight of the nine cycles performed. For the final flow reduction, the mass flux was decreased to 6.1 x 10/sup 5/ lb/hr-ft/sup 2/ at a constant peak linear heat generation rate of 17.8 kW/ft. The fuel rod was allowed to remain in film boiling for about 210 seconds during this final flow reduction. The fuel rod remained intact during the test. Results of on-line measurements of the fuel rod behavior are presented together with discussion of instrument performance. A comparison of the data with Fuel Rod Analysis Program-Transient 2 (FRAP-T2) computer code calculations is included.

  16. Addressing the Axial Burnup Distribution in PWR Burnup Credit Criticality Safety

    International Nuclear Information System (INIS)

    This paper summarizes efforts related to developing a technically justifiable approach for addressing the axial burnup distribution in PWR burnup-credit criticality safety analyses. The paper reviews available data on the axial variation in burnup and the effect of axial burnup profiles on reactivity in a SNF cask. A publicly available database of profiles is examined to identify profiles that maximize the neutron multiplication factor, keff, assess its adequacy for general PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. For this assessment, a statistical evaluation of the keff values associated with the profiles in the axial burnup profile database was performed that identifies the most reactive profiles as statistical outliers that are not representative of typical discharged SNF assemblies. The impact of these bounding profiles on the neutron multiplication factor for a high-density burnup credit cask is quantified. Finally, analyses are presented to quantify the potential reactivity consequence of assemblies with axial profiles that are not bounded by the existing database. The paper concludes with findings for addressing the axial burnup distribution in burnup credit analyses

  17. Survey of prediction capabilities of three nuclear data libraries for a PWR application

    International Nuclear Information System (INIS)

    Highlights: • State of uncertainty quantification of three NDLs for a PWR application. • Identification of several order-of-magnitude differences in important contributors. • Impact of cross-material correlations. - Abstract: We survey prediction capabilities of ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0u nuclear data libraries (NDLs) for the application of generating two-group homogenized assembly constants for a steady state diffusion model in the context of UAM-LWR (Uncertainty Analysis in Best-Estimate Modeling for Design, Operation and Safety Analysis of LWRs) Benchmark. We consider two different fuel assembly test cases representing a PWR. State of uncertainty quantification in each NDLs is presented for the application. We expect small differences between the NDLs due to the use of expert judgment in the evaluation processes, and identify several order-of-magnitude differences between the NDLs for significant contributors to uncertainty. We also quantify the contribution from cross-material correlations to the uncertainties

  18. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    International Nuclear Information System (INIS)

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety

  19. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  20. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

    2015-02-15

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  1. Improvement of the fracture toughness in the PWR pressurizer surgeline material (Final report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Oh, Yong Jun; Yoon, Ji Hyun; Oh, Jong Myung; Park, Buk Gyun; Kim, Ju Hak; Kuk Il Hyun; Byun, Taek Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Fracture toughness property of PWR surgeline pipe materials is one of the most important factor for Leak Before Break(LBB) analysis. In order to improve fracture toughness of the surgeline material (SA312-TP347 stainless steel), base on the evaluation and analysis of the commercial TP347 alloys, eleven TP347 model alloys were designed and manufactured. Tensile and fracture resistance properties of the model alloys, as well as microstructure, were evaluated. It is concluded that the nitrogen shall be added more than 0.1% for high tensile property and the carbon shall be in the range of 0.02 to 0.04% for high fracture resistance. In addition, four TP316N stainless steels were manufactured and evaluated to find out the applicability as a candidate material for PWR surgeline pipe. As a conclusion, TP316N stainless steels had an excellent property to be used for surgeline piping materials, substituting the present TP347 stainless steels. (author). 11 refs., 50 figs., 12 tabs.

  2. Verification test for radiation reduction effect and material integrity on PWR primary system by zinc injection

    International Nuclear Information System (INIS)

    Zinc injection is known to be an effective method for the reduction of radiation source in the primary water system of a PWR. There is a need to verify the effect of Zn injection operation on radiation source reduction and materials integrity of PWR primary circuit. In order to confirm the effectiveness of Zn injection, verification test as a national program sponsored by Ministry of Economy, Trade and Industry (METI) was started in 1995 for 7-year program, and will be finished by the end of March in 2002. This program consists of irradiation test and material integrity test. Irradiation test as an In-Pile-Test managed by AEAT Plc(UK) was performed using the LVR-15 reactor of NRI Rez in Check Republic. Furthermore, Out-of-Pile-Test using film adding unit was also performed to obtain supplemental data for In-Pile-Test at Takasago Engineering Laboratory of NUPEC. Material Integrity test was planned to perform constant load test, constant strain test and corrosion test at the same time using large scale Loop and slow strain extension rate testing (SSRT) at Takasago Engineering Laboratory of NUPEC. In this paper, the results of the verification test for Zinc program at present are discussed. (authors)

  3. A quantitative comparison of loading pattern optimization methods for in-core fuel management of PWR

    International Nuclear Information System (INIS)

    The performance of several loading pattern(LP) optimization methods was quantitatively compared through a benchmark problem of PWR LP optimization. The simulated annealing(SA) method, the genetic algorithms(GA) method, the direct search(DS) method based on assembly multiple shuffling and the binary exchange(BE) method based on fuel assembly binary exchange were investigated as candidates for the optimization techniques. Hybrid strategy which combined different optimization methods was newly proposed, and the performances of two different new hybrid methods, which combined DS with BE and GA with BE were examined. From the results of the LP optimization benchmark problem, the superiority and inferiority of each method were clarified. Furthermore, it was demonstrated that the GA+BE hybrid strategy performed best among these methods. By combining GA with BE, the weaknesses of these two methods were compensated for with each other and the optimization performance was improved significantly. Therefore, the GA+BE hybrid method is quite effective for the LP optimization problems of PWR. (author)

  4. Analytical one-dimensional frequency response and stability model for PWR nuclear power plants

    International Nuclear Information System (INIS)

    A dynamic model for PWR nuclear power plants is presented. The plant is assumed to consist of one-dimensional single-channel core, a counterflow once-through steam generator (represented by two nodes according to the nonboiling and boiling region) and the necessary connection coolant lines. The model describes analytically the frequency response behaviour of important parameters of such a plant with respect to perturbations in reactivity, subcooling or mass flow (both at the entrances to the reactor core and/or the secondary steam generator side), the perturbations in steam load or system pressure (on the secondary side of the steam generator). From corresponding 'open' loop considerations it can then be concluded - by applying the Nyquist criterion - upon the degree of the stability behaviour of the underlying system. Based on this theoretical model, a computer code named ADYPMO has been established. From the knowledge of the frequency response behaviour of such a system, the corresponding transient behaviour with respect to a stepwise or any other perturbation signal can also be calculated by applying an appropriate retransformation method, e.g. by using digital code FRETI. To demonstrate this procedure, a transient experimental curve measured during the pre-operational test period at the PWR nuclear power plant KKS Stade was recalculated using the combination ADYPMO-FRETI. Good agreement between theoretical calculations and experimental results give an insight into the validity and efficiency of the underlying theoretical model and the applied retransformation method. (Auth.)

  5. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    International Nuclear Information System (INIS)

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs

  6. Modeling of the thermo-mechanical behaviour of the PWR fuel

    International Nuclear Information System (INIS)

    This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

  7. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  8. Cytotoxic and genotoxic evaluation of orthodontic adhesives with primer and without primer exposed to electron beam irradiation - an in-vitro study

    International Nuclear Information System (INIS)

    To evaluate the in vitro genotoxicity and cytotoxicity of two visible light-cured adhesives. The materials tested were 1. orthodontic adhesive with primer (Transbond XT3M) and 2. Orthodontic adhesive without primer (Heliosit, Ivoclar Vivadent AG), Cured sterile individual masses were exposed to 2 kGy electron beam radiation, both irradiated and non irradiated materials were immersed in Phosphate buffer saline and left at 370℃ for 24 hr. Then a volume of 200 μL of the extract medium was mixed with human peripheral blood lymphocyte tested for comet assay by single cell DNA Damage assay and Apoptosis by DNA diffusion agar assay. Evaluation of cytotoxicity was carried out by Hemolysis assay method. Haemolytic activity of orthodontic adhesive without primer (53.34±3.12) was slightly more than that of orthodontic adhesive with primer (52.9±.88). In case of Apoptosis, adhesive with primer (188.92±55.05) and adhesives without primer (186.75±101.83) showed increased diffusion of DNA compared to normal lymphocyte (111.22±8.78). However the level of DNA diffusion was not significantly different between the two adhesives. Both adhesives were cytotoxic and induced apoptosis. Adhesives without primer were found to be slightly toxic than that of adhesive with primer. Both the adhesives had no significant effect on the percentage of DNA tail and olive tail moment of DNA exposed to electron beam radiation. (author)

  9. AP1000® PWR reactor physics analysis with VERA-CS and KENO-VI. Part 2. Power distribution

    International Nuclear Information System (INIS)

    Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000® PWR. The AP1000 PWR features an advanced first core with radial and axial heterogeneities, including enrichment zoning, multiple burnable absorbers, and a combination of light and heavy control banks to enable the MSHIMTM advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. A companion paper at this conference describes the results obtained with VERA-CS and the KENO Monte-Carlo code for startup physics tests simulations of the AP1000 PWR first core (critical boron, rod worth and reactivity coefficients). This paper describes the results of detailed power distribution comparisons between VERA-CS and KENO, and confirms the excellent numerical agreement reported in the companion paper for the startup physics tests simulations. (author)

  10. Diagnosis and prognosis of the source term by the French Safety Institut during an emergency on a PWR

    International Nuclear Information System (INIS)

    The French approach for the diagnosis and the prognosis of the source term during an accident on a PWR is presented and the tools which have been developed to implement this approach at the Institute for Nuclear Protection and Safety (IPSN) are described. (author). 2 refs, 3 figs

  11. Research on General Corrosion Property of 304L and 304NG Stainless Steels in Simulated PWR Primary Water

    Institute of Scientific and Technical Information of China (English)

    PENG; De-quan; HU; Shi-lin; ZHANG; Ping-zhu; WANG; Hui

    2012-01-01

    <正>The general corrosion behaviors of 304L and 304NG grade stainless steels in simulated pressurized water reactor (PWR) primary loop were studied using still autoclave, respectively, the corrosion test lasted for 1 680 hours. The corrosion oxide films were analyzed macroscopically and microscopically. The results are shown in Figs. 1, 2.

  12. Influence of localized deformation on A-286 austenitic stainless steel stress corrosion cracking in PWR primary water

    International Nuclear Information System (INIS)

    Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels is known to be a critical issue for structural components of nuclear reactor cores. The deformation of irradiated austenitic stainless steels is extremely heterogeneous and localized in deformation bands that may play a significant role in IASCC. In this study, an original approach is proposed to determine the influence of localized deformation on austenitic stainless steels SCC in simulated PWR primary water. The approach consists in (i) performing low cycle fatigue tests on austenitic stainless steel A-286 strengthened by γ' precipitates Ni3(Ti,Al) in order to shear and dissolve the precipitates in intense slip bands, leading to a localization of the deformation within and in (ii) assessing the influence of these γ'-free localized deformation bands on A-286 SCC by means of comparative CERT tests performed on specimens with similar yield strength, containing or not γ'-free localized deformation bands. Results show that strain localization significantly promotes A-286 SCC in simulated PWR primary water at 320 and 360 C. Moreover, A-286 is a precipitation-hardening austenitic stainless steel used for applications in light water reactors. The second objective of this work is to gain insights into the influence of heat treatment and metallurgical structure on A-286 SCC susceptibility in PWR primary water. The results obtained demonstrate a strong correlation between yield strength and SCC susceptibility of A-286 in PWR primary water at 320 and 360 C. (author)

  13. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation

  14. Bt-CoV HKU3 PCR/Sequencing Primers

    OpenAIRE

    sprotocols

    2014-01-01

    Authors: Rachel Graham ### Abstract This protocol details the primers and conditions used for forward and reverse PCR amplification and sequencing of Bt-CoV HKU3 genomes. ### Introduction This protocol details the steps, reagents, and conditions required to sequence Bt-CoV HKU3 genomes in the forward and reverse directions. The protocol begins with the RT-PCR step, assuming that BtCoV HKU3 RNA purified using a standard procedure (i.e., TRIzol extraction) has already been perfo...

  15. Bt-CoV HKU5 PCR/Sequencing Primers

    OpenAIRE

    sprotocols

    2014-01-01

    Authors: Rachel Graham ### Abstract This protocol details the primers and conditions used for forward and reverse PCR amplification and sequencing of Bt-CoV HKU5 genomes. ### Introduction This protocol details the steps, reagents, and conditions required to sequence Bt-CoV HKU5 genomes in the forward and reverse directions. The protocol begins with the RT-PCR step, assuming that BtCoV HKU5 RNA purified using a standard procedure (i.e., TRIzol extraction) has already been perfo...

  16. Accelerating MATLAB with GPU computing a primer with examples

    CERN Document Server

    Suh, Jung W

    2013-01-01

    Beyond simulation and algorithm development, many developers increasingly use MATLAB even for product deployment in computationally heavy fields. This often demands that MATLAB codes run faster by leveraging the distributed parallelism of Graphics Processing Units (GPUs). While MATLAB successfully provides high-level functions as a simulation tool for rapid prototyping, the underlying details and knowledge needed for utilizing GPUs make MATLAB users hesitate to step into it. Accelerating MATLAB with GPUs offers a primer on bridging this gap. Starting with the basics, setting up MATLAB for

  17. A finite element primer for beginners the basics

    CERN Document Server

    Zohdi, Tarek I

    2014-01-01

    The purpose of this primer is to provide the basics of the Finite Element Method, primarily illustrated through a classical model problem, linearized elasticity. The topics covered are:(1) Weighted residual methods and Galerkin approximations,(2) A model problem for one-dimensional?linear elastostatics,(3) Weak formulations in one dimension,(4) Minimum principles in one dimension,(5) Error estimation in one dimension,(5) Construction of Finite Element basis functions in one dimension,(6) Gaussian Quadrature,(7) Iterative solvers and element by element data structures,(8) A model problem for th

  18. Guidelines - A Primer for Communicating Effectively with NABIR Stakeholders

    Energy Technology Data Exchange (ETDEWEB)

    Weber, James R.; Schell, Charlotte J.; Marino, T; Bilyard, Gordon R.

    2004-02-10

    This version of the communication primer comprises two interlocking parts: Pat 1, a practical section, intended to prepare you for public interactions, and Part 2, a theoretical section that provides social and technical bases for the practices recommended in Part 1. The mutual support of practice and theory is very familiar in science and clearly requires a willingness to observe and revise our prior assumptions--in this document, we invoke both. We hope that is offering will represent a step both towards improving practice and maturing the theory of practical science communication.

  19. Atmel AVR Microcontroller Primer Programming and Interfacing, Second Edition

    CERN Document Server

    Barrett, Steven F

    2012-01-01

    This textbook provides practicing scientists and engineers a primer on the Atmel AVR microcontroller. In this second edition we highlight the popular ATmega164 microcontroller and other pin-for-pin controllers in the family with a complement of flash memory up to 128 kbytes. The second edition also adds a chapter on embedded system design fundamentals and provides extended examples on two different autonomous robots. Our approach is to provide the fundamental skills to quickly get up and operating with this internationally popular microcontroller. We cover the main subsystems aboard the ATmega

  20. Automated guided vehicle systems a primer with practical applications

    CERN Document Server

    Ullrich, Günter

    2015-01-01

    This primer is directed at experts and practitioners in intralogistics who are concerned with optimizing material flows. The presentation is comprehensive covering both, practical and theoretical aspects with a moderate degree of specialization, using clear and concise language. Areas of operation as well as technical standards of all relevant components and functions are described. Recent developments in technology and in the markets are taken into account. The goal of this book is to further stronger use of automated guided transport systems and the enhancement of their future performance.

  1. Transport phenomena in Newtonian fluids a concise primer

    CERN Document Server

    Olsson, Per

    2013-01-01

    This short primer provides a concise and tutorial-style introduction to transport phenomena in Newtonian fluids , in particular the transport of mass, energy and momentum.  The reader will find detailed derivations of the transport equations for these phenomena, as well as selected analytical solutions to the transport equations in some simple geometries. After a brief introduction to the basic mathematics used in the text, Chapter 2, which deals with momentum transport, presents a derivation of the Navier-Stokes-Duhem equation describing the basic flow in a Newtonian fluid.  Also provided at

  2. Steel Primer Chamber Assemblies for Dual Initiated Pyrovalves

    Science.gov (United States)

    Guemsey, Carl S.; Mizukami, Masashi; Zenz, Zac; Pender, Adam A.

    2009-01-01

    A solution was developed to mitigate the potential risk of ignition failures and burn-through in aluminum primer chamber assemblies on pyrovalves. This was accomplished by changing the assembly material from aluminum to steel, and reconfiguration of flame channels to provide more direct paths from initiators to boosters. With the geometric configuration of the channels changed, energy is more efficiently transferred from the initiators to the boosters. With the alloy change to steel, the initiator flame channels do not erode upon firing, eliminating the possibility of burn-through. Flight qualification tests have been successfully passed.

  3. A primer on physical-layer network coding

    CERN Document Server

    Liew, Soung Chang; Zhang, Shengli

    2015-01-01

    The concept of physical-layer network coding (PNC) was proposed in 2006 for application in wireless networks. Since then it has developed into a subfield of communications and networking with a wide following. This book is a primer on PNC. It is the outcome of a set of lecture notes for a course for beginning graduate students at The Chinese University of Hong Kong. The target audience is expected to have some prior background knowledge in communication theory and wireless communications, but not working knowledge at the research level. Indeed, a goal of this book/course is to allow the reader

  4. The primer for sports medicine professionals on imaging: the shoulder.

    Science.gov (United States)

    Farshad-Amacker, Nadja A; Jain Palrecha, Sapna; Farshad, Mazda

    2013-01-01

    Because of its inherent superior soft tissue contrast and lack of ionizing radiation, magnetic resonance imaging (MRI) is highly suited to study the complex anatomy of the shoulder joint, particularly when assessing the relatively high incidence of shoulder injuries in young, athletic patients. This review aims to serve as a primer for understanding shoulder MRI in an algorithmical approach, including MRI protocol and technique, normal anatomy and anatomical variations of the shoulder, pathologic conditions of the rotator cuff tendons and muscles, the long head of the biceps tendon, shoulder impingement, labral and glenohumeral ligament pathology, MR findings in shoulder instability, adhesive capsulitis, and osteoarthritis. PMID:24381700

  5. Discrete random signal processing and filtering primer with Matlab

    CERN Document Server

    Poularikas, Alexander D

    2013-01-01

    Engineers in all fields will appreciate a practical guide that combines several new effective MATLAB® problem-solving approaches and the very latest in discrete random signal processing and filtering.Numerous Useful Examples, Problems, and Solutions - An Extensive and Powerful ReviewWritten for practicing engineers seeking to strengthen their practical grasp of random signal processing, Discrete Random Signal Processing and Filtering Primer with MATLAB provides the opportunity to doubly enhance their skills. The author, a leading expert in the field of electrical and computer engineering, offe

  6. A plastome primer set for comprehensive quantitative real time RT-PCR analysis of Zea mays: a starter primer set for other Poaceae species

    Directory of Open Access Journals (Sweden)

    Dunn Sade N

    2008-06-01

    Full Text Available Abstract Background Quantitative Real Time RT-PCR (q2(RTPCR is a maturing technique which gives researchers the ability to quantify and compare very small amounts of nucleic acids. Primer design and optimization is an essential yet time consuming aspect of using q2(RTPCR. In this paper we describe the design and empirical optimization of primers to amplify and quantify plastid RNAs from Zea mays that are robust enough to use with other closely related species. Results Primers were designed and successfully optimized for 57 of the 104 reported genes in the maize plastome plus two nuclear genes. All 59 primer pairs produced single amplicons after end-point reverse transcriptase polymerase chain reactions (RT-PCR as visualized on agarose gels and subsequently verified by q2(RTPCR. Primer pairs were divided into several categories based on the optimization requirements or the uniqueness of the target gene. An in silico test suggested the majority of the primer sets should work with other members of the Poaceae family. An in vitro test of the primer set on two unsequenced species (Panicum virgatum and Miscanthus sinensis supported this assumption by successfully producing single amplicons for each primer pair. Conclusion Due to the highly conserved chloroplast genome in plant families it is possible to utilize primer pairs designed against one genomic sequence to detect the presence and abundance of plastid genes or transcripts from genomes that have yet to be sequenced. Analysis of steady state transcription of vital system genes is a necessary requirement to comprehensively elucidate gene expression in any organism. The primer pairs reported in this paper were designed for q2(RTPCR of maize chloroplast genes but should be useful for other members of the Poaceae family. Both in silico and in vitro data are presented to support this assumption.

  7. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    International Nuclear Information System (INIS)

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50

  8. Image-Processing-Based Study of the Interfacial Behavior of the Countercurrent Gas-Liquid Two-Phase Flow in a Hot Leg of a PWR

    Directory of Open Access Journals (Sweden)

    Gustavo A. Montoya

    2012-01-01

    Full Text Available The interfacial behavior during countercurrent two-phase flow of air-water and steam-water in a model of a PWR hot leg was studied quantitatively using digital image processing of a subsequent recorded video images of the experimental series obtained from the TOPFLOW facility, Helmholtz-Zentrum Dresden-Rossendorf e.V. (HZDR, Dresden, Germany. The developed image processing technique provides the transient data of water level inside the hot leg channel up to flooding condition. In this technique, the filters such as median and Gaussian were used to eliminate the drops and the bubbles from the interface and the wall of the test section. A Statistical treatment (average, standard deviation, and probability distribution function (PDF of the obtained water level data was carried out also to identify the flow behaviors. The obtained data are characterized by a high resolution in space and time, which makes them suitable for the development and validation of CFD-grade closure models, for example, for two-fluid model. This information is essential also for the development of mechanistic modeling on the relating phenomenon. It was clarified that the local water level at the crest of the hydraulic jump is strongly affected by the liquid properties.

  9. BEACON/MOD2A: a computer program for subcompartment analysis of nuclear reactor containment. A user's manual. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Wells, R. A.

    1979-03-01

    The BEACON code is a Best Estimate Advanced Containment code which being developed by EG and G, Idaho, Inc., at the Idaho National Engineering Laboratory. The program is designed to perform a best estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system. The current version of BEACON, which is designated BEACON/MOD2A, contains mass and heat transfer models for wall film and for wall conduction. It is suitable for the evaluation of short term transients in PWR dry containment systems. This manual describes the models employed in BEACON/MOD2A and specifies code implementation requirements. It provides application information for input data preparation and for output data interpretation.

  10. Effect of cavity preparation method on microtensile bond strength of a self-etching primer vs phosphoric acid etchant to enamel.

    Science.gov (United States)

    de Souza-Zaroni, Wanessa Christine; Delfino, Carina Sinclér; Ciccone-Nogueira, Juliane Cristina; Palma-Dibb, Regina Guenka; Corona, Silmara Aparecida Milori

    2007-10-01

    This study evaluated the effect of cavity preparation using air abrasion or carbide bur on bond strength to enamel treated with a self-etching primer (Tyrian SPE) or a phosphoric acid etchant. Twenty-four molars were divided into three groups: high-speed; standard handpiece (ST air abrasion) or supersonic handpiece (SP air abrasion) of the same air-abrasive system. The enamel surfaces were treated with one of the two etchants and the same adhesive agent One Step Plus, and then composite buildups were done with Filtek Z250. After 24 h at 37 degrees C, beams (0.8 mm2) were obtained and subjected to tensile stress in a universal testing machine (0.5 mm/min). The data were submitted to analysis of variance and Tukey's test (P cavity preparation method was dependent on the conditioning system used, only when using carbide-bur preparation technique.

  11. Les primeres locutores de ràdio a Catalunya, dels inicis de l’invent al primer franquisme

    OpenAIRE

    Espinosa i Mirabet, Sílvia

    2009-01-01

    Aquest article posa de manifest el paper preponderant que tingueren les locutores en les emissores de ràdio que operaven a Catalunya sota l’epígraf EAJ durant els anys vint i trenta del segle XX. Les locutores foren les primeres professionals del mitjà que crearen programes (els femenins) amb un format que més endavant es coneixeria com a magazín, en un moment històric (anys vint) en el qual la ràdio estava plena de conferències i xerrades. La segona gran aportació de les dones a la història ...

  12. Microsatellite Primer Development for Post Oak, Quercus stellata (Fagaceae

    Directory of Open Access Journals (Sweden)

    Warren B. Chatwin

    2014-10-01

    Full Text Available Premise of the study: The American Cross Timbers forest ecosystem runs from southeastern Kansas to Central Texas and is primarily composed of post oak (Quercus stellata. This old-growth forest currently occupies only about 2% of its ancestral range. To facilitate genetic research on this species, we developed microsatellite primers specific to post oak from reduced genomic libraries. Methods and Results: Two Q. stellata individuals, sampled from the northern and southern range of the post oak forest, were subject to genomic reduction and 454 pyrosequencing. Bioinformatic analysis identified putative microsatellites from which 12 polymorphic primer sets were screened on three populations. The number of alleles observed ranged from five to 20 across all populations, while observed and expected heterozygosity values ranged from 0.05 to 0.833 and 0.236 to 0.893, respectively, within individual populations. Conclusions: We report the development of microsatellite markers, specific to post oak, to aid the study of genetic diversity and population structure of extant forest remnants.

  13. Primer for evaluating ecological risk at petroleum release sites.

    Science.gov (United States)

    Claff, R

    1999-02-01

    Increasingly, risk-based approaches are being used to guide decision making at sites such as service stations and petroleum product terminals, where petroleum products have been inadvertently released to the soil. For example, the API Decision Support System software, DSS, evaluates site human health risk along six different routes of exposure. The American Society for Testing and Materials' Risk-Based Corrective Action (RBCA) standard, ASTM 1739, establishes a tiered framework for evaluating petroleum release sites on the basis of human health risk. Though much of the risk assessment focus has been on human health risk, regulatory agencies recognize that protection of human health may not fully protect the environment; and EPA has developed guidance on identifying ecological resources to be protected through risk-based decision making. Not every service station or petroleum product terminal site warrants a detailed ecological risk assessment. In some cases, a simple preliminary assessment will provide sufficient information for decision making. Accordingly, the American Petroleum Institute (API) is developing a primer for site managers, to assist them in conducting this preliminary assessment, and in deciding whether more detailed ecological risk assessments are warranted. The primer assists the site manager in identifying relevant ecological receptors and habitats, in identifying chemicals and exposure pathways of concern, in developing a conceptual model of the site to guide subsequent actions, and in identifying conditions that may warrant immediate response. PMID:10189585

  14. FullSSR: Microsatellite Finder and Primer Designer.

    Science.gov (United States)

    Metz, Sebastián; Cabrera, Juan Manuel; Rueda, Eva; Giri, Federico; Amavet, Patricia

    2016-01-01

    Microsatellites are genomic sequences comprised of tandem repeats of short nucleotide motifs widely used as molecular markers in population genetics. FullSSR is a new bioinformatic tool for microsatellite (SSR) loci detection and primer design using genomic data from NGS assay. The software was tested with 2000 sequences of Oryza sativa shotgun sequencing project from the National Center of Biotechnology Information Trace Archive and with partial genome sequencing with ROCHE 454® from Caiman latirostris, Salvator merianae, Aegla platensis, and Zilchiopsis collastinensis. FullSSR performance was compared against other similar SSR search programs. The results of the use of this kind of approach depend on the parameters set by the user. In addition, results can be affected by the analyzed sequences because of differences among the genomes. FullSSR simplifies the detection of SSRs and primer design on a big data set. The command line interface of FullSSR was intended to be used as part of genomic analysis tools pipeline; however, it can be used as a stand-alone program because the results are easily interpreted for a nonexpert user. PMID:27366148

  15. FullSSR: Microsatellite Finder and Primer Designer

    Directory of Open Access Journals (Sweden)

    Sebastián Metz

    2016-01-01

    Full Text Available Microsatellites are genomic sequences comprised of tandem repeats of short nucleotide motifs widely used as molecular markers in population genetics. FullSSR is a new bioinformatic tool for microsatellite (SSR loci detection and primer design using genomic data from NGS assay. The software was tested with 2000 sequences of Oryza sativa shotgun sequencing project from the National Center of Biotechnology Information Trace Archive and with partial genome sequencing with ROCHE 454® from Caiman latirostris, Salvator merianae, Aegla platensis, and Zilchiopsis collastinensis. FullSSR performance was compared against other similar SSR search programs. The results of the use of this kind of approach depend on the parameters set by the user. In addition, results can be affected by the analyzed sequences because of differences among the genomes. FullSSR simplifies the detection of SSRs and primer design on a big data set. The command line interface of FullSSR was intended to be used as part of genomic analysis tools pipeline; however, it can be used as a stand-alone program because the results are easily interpreted for a nonexpert user.

  16. A passive decay heat removal system for LWRs based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2015-05-15

    Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

  17. Selection of functional tRNA primers and primer binding site sequences from a retroviral combinatorial library: identification of new functional tRNA primers in murine leukemia virus replication

    DEFF Research Database (Denmark)

    Lund, Anders Henrik; Duch, M; Pedersen, F S

    2000-01-01

    Retroviral reverse transcription is initiated from a cellular tRNA molecule and all known exogenous isolates of murine leukemia virus utilise a tRNA(Pro)molecule. While several studies suggest flexibility in murine leukemia virus primer utilisation, studies on human immunodeficiency virus and avian...... retro-viruses have revealed evidence of molecular adapt-ation towards the specific tRNA isoacceptor used as replication primer. In this study, murine leukemia virus tRNA utilisation is investigated by in vivo screening of a retroviral vector combinatorial library with randomised primer binding sites....... While most of the selected primer binding sites are complementary to the 3'-end of tRNA((Pro)), we also retrieved PBS sequences matching four other tRNA molecules and demonstrate that Akv murine leukemia virus vectors may efficiently replicate using tRNA(Arg(CCU)), tRNA(Phe(GAA))and a hitherto unknown...

  18. Performance testing of lead free primers: blast waves, velocity variations, and environmental testing

    CERN Document Server

    Courtney, Elya; Summer, Peter David; Courtney, Michael

    2014-01-01

    Results are presented for lead free primers based on diazodinitrophenol (DDNP)compared with tests on lead styphnate based primers. First, barrel friction measurements in 5.56 mm NATO are presented. Second, shot to shot variations in blast waves are presented as determined by detonating primers in a 7.62x51mm rifle chamber with a firing pin, but without any powder or bullet loaded and measuring the blast wave at the muzzle with a high speed pressure transducer. Third, variations in primer blast waves, muzzle velocities, and ignition delay are presented after environmental conditioning (150 days) for two lead based and two DDNP based primers under cold and dry (-25 deg C,0% relative humidity), ambient (20 deg C, 50% relative humidity), and hot & humid (50 deg C, 100% relative humidity) conditions in 5.56 mm NATO. Taken together, these results indicate that DDNP based primers are not sufficiently reliable for service use.

  19. A sputtered zirconia primer for improved thermal shock resistance of plasma-sprayed ceramic turbine seals

    Science.gov (United States)

    Bill, R. C.; Sovey, J.; Allen, G. P.

    1981-01-01

    It is shown that the application of sputtered Y2O3-stabilized ZrO2 (YSZ) primer in plasma-sprayed YSZ ceramic-coated turbine blades results in an improvement, by a factor of 5-6, in the thermal shock life of specimens with a sprayed, porous, Ni-Cr-Al-Y intermediate layer. Species with and without the primer were found to be able to survive 1000 cycles when the intermediate layer was used, but reduced laminar cracking was observed in the specimen with the primer. It is suggested that the sputtered YZS primer-induced properties are due to (1) more effective wetting and adherence of the plasma-sprayed YZS particles to the primer, and (2) the primer's retardation of impinging, molten plasma sprayed particles solidification rates, which result in a less detrimental residual stress distribution.

  20. Specific and sensitive quantitative RT-PCR of miRNAs with DNA primers

    DEFF Research Database (Denmark)

    Balcells, Ingrid; Cirera Salicio, Susanna; Busk, Peter K.

    2011-01-01

    be designed with a success rate of 94%. The method was able to quantify synthetic templates over eight orders of magnitude and readily discriminated between microRNAs with single nucleotide differences. Importantly, PCR with DNA primers yielded significantly higher amplification efficiencies of biological...... samples than a similar method based on locked nucleic acids-spiked primers, which is in agreement with the observation that locked nucleic acid interferes with efficient amplification of short templates. The higher amplification efficiency of DNA primers translates into higher sensitivity and precision...... settings. RESULTS: We describe a PCR method for quantification of microRNAs based on a single reverse transcription reaction for all microRNAs combined with real-time PCR with two, microRNA-specific DNA primers. Primer annealing temperatures were optimized by adding a DNA tail to the primers and could...