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Sample records for air primer pwr

  1. Air Pollution Primer.

    Science.gov (United States)

    National Tuberculosis and Respiratory Disease Association, New York, NY.

    As the dangers of polluted air to the health and welfare of all individuals became increasingly evident and as the complexity of the causes made responsibility for solutions even more difficult to fix, the National Tuberculosis and Respiratory Disease Association felt obligated to give greater emphasis to its clean air program. To this end they…

  2. Evaluation of a PWR assembly subjected to air-storage

    International Nuclear Information System (INIS)

    Most dry storage tests and demonstrations involving Zircaloy-clad fuel have been conducted in inert cover gases. The availability of irradiated PWR fuel and an instrumented dry storage test module (Fuel Temperature Test or FTT Facility) at the Engine Maintenance and Disassembly (EMAD) site in Nevada prompted a dry storage test in air. Not included in the prior publication are results of the post-test fuel examination. To date, that examination has involved visual inspection, photography, and analysis of swipes taken from fuel assembly surfaces. This paper summarizes results of the examination and an overview of the test interpretation and significance

  3. 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.

  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. 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)

  6. Cocurrent flow in an atmospheric pressure air-water model of a PWR hot leg at 0.114 scale

    International Nuclear Information System (INIS)

    Experimental work has been carried out with air and water in a 0.114 times scale model of the proposed Sizewell 'B' PWR hot leg. The model was accurate from the entry from the reactor vessel to the tube plate in the steam generator. Visual observations are summarized in a flow map in the coordinates of an air Froude number versus a water Froude number. The map shows a predominantly stratified system with spray occurring at higher air flowrates and surges appearing at higher water flowrates. Also, at very low air flowrates, long bubbles of air passed along the top of the hot leg through water filling the pipe. A previous theory predicts the water level in the reactor vessel in terms of the air and water flowrates. This theory also agrees with data from other sources on the level in the hot leg in terms of the discharge of two phases through a horizontal hot leg break. The data includes results for steam and water at pressures up to 62 bar. Correlations for the interface level in the hot leg are developed. Data for lower air flowrates agrees substantially with a theory for the onset of slugging. (author)

  7. Air-water model studies of cocurrent flow into and along a PWR hot leg to the steam generator

    International Nuclear Information System (INIS)

    An experimental study is made of the cocurrent flow of air and water at atmospheric pressure from a reactor vessel into and along an approximately ninth-scale replica of the Sizewell 'B' PWR hot leg to the steam generator. A flow regime map of conditions in the hot leg is presented. The water interface level in the reactor vessel as a function of the flowrates is in agreement with a recent theory developed by the author. The same theory predicts the level in the hot leg when discharging two phases through a horizontal break and is in agreement with the results of other workers on this subject for the discharge of air and water up to a pressure of 5 bar and of steam and water up to a pressure of 62 bar. Results on the water level in the hot leg are correlated empirically but, for lower flowrates, the results are in approximate agreement with a theory for the onset of flooding. (orig.)

  8. Measurement of pressure drops in prototypic BWR and PWR fuel assemblies in the laminar regime - Pressure drop measurement of laminar air flow in prototypic BWR and PWR fuel assemblies

    International Nuclear Information System (INIS)

    Laminar pressure drops in nuclear fuel assemblies are of interest for evaluating complete loss-of-coolant accident scenarios in spent fuel pools and for performance analyses of dry storage casks. To the knowledge of the authors, this study represents the first attempt to directly quantify pressure losses in prototypic fuel assemblies in the laminar regime. Two commercial fuel assemblies were examined including a 17x17 PWR and a 9x9 BWR. The assemblies were tested in the laminar regime with Reynolds numbers ranging from 10 to 1000, based on the average assembly velocity and hydraulic diameter. Pressure drop measurements were made across individual bundle spans and grid spacers in the mock fuel assemblies using high-sensitivity differential pressure gauges. These gauges are capable of detecting extremely small changes in differential pressure with a resolution of ∼0.02 Pa. This level of sensitivity allows meaningful pressure drop measurements across separate fuel components, even at low Reynolds numbers. The fuel assembly mock-ups were constructed from commercial fuel assembly structural components and stainless steel tubing that is within 0.6 pc of the outer diameter of actual fuel. The outer flow boundary in the BWR assembly bundle was defined by the walls of a prototypic canister. In the PWR assembly, the flow was confined by the walls of different stainless steel storage cells. Two of the PWR storage cell sizes represented dimensions spanning pool and cask cells available in industry. Pressure ports were installed along the length of the assemblies at locations corresponding to the entrance and exit of fuel components. Dry, ambient air was metered into the bottom of each assembly through a flow straightener. The geometries of the tube bundles in 17x17 PWR and 9x9 BWR fuel assemblies are fundamentally different. The PWR bundle has a larger flow area and incorporates more grid spacers compared to the BWR bundle. Additionally, eight of the 74 fuel rods in the 9x9

  9. Air--water countercurrent annular flow in vertical tubes. Interim report. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bharathan, D.

    1978-05-01

    Air--water countercurrent flow characteristics in 2.5 and 5.1 cm vertical tubes are investigated. Experimental measurements include air and water flow rates, pressure losses, pressure gradients, and liquid fractions. Tube-end geometries are altered to study their influence on the flow characteristics. Liquid-fraction measurements indicate that the countercurrent flow may be divided into three regions based upon the relative magnitudes of interfacial and wall shear stresses. The dependence of interfacial friction factor on the liquid fraction is isolated. The mechanism limiting countercurrent flows within a tube is modelled by a simple theory. Salient features of the theory are demonstrated. Comparisons between the theory and some experimental data are presented.

  10. Network Theory: A Primer and Questions for Air Transportation Systems Applications

    Science.gov (United States)

    Holmes, Bruce J.

    2004-01-01

    A new understanding (with potential applications to air transportation systems) has emerged in the past five years in the scientific field of networks. This development emerges in large part because we now have a new laboratory for developing theories about complex networks: The Internet. The premise of this new understanding is that most complex networks of interest, both of nature and of human contrivance, exhibit a fundamentally different behavior than thought for over two hundred years under classical graph theory. Classical theory held that networks exhibited random behavior, characterized by normal, (e.g., Gaussian or Poisson) degree distributions of the connectivity between nodes by links. The new understanding turns this idea on its head: networks of interest exhibit scale-free (or small world) degree distributions of connectivity, characterized by power law distributions. The implications of scale-free behavior for air transportation systems include the potential that some behaviors of complex system architectures might be analyzed through relatively simple approximations of local elements of the system. For air transportation applications, this presentation proposes a framework for constructing topologies (architectures) that represent the relationships between mobility, flight operations, aircraft requirements, and airspace capacity, and the related externalities in airspace procedures and architectures. The proposed architectures or topologies may serve as a framework for posing comparative and combinative analyses of performance, cost, security, environmental, and related metrics.

  11. 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.

  12. 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…

  13. Investigations on the aging of activated carbons in the exhaust air of a pressurized water reactor (PWR 4)

    International Nuclear Information System (INIS)

    Investigations were performed on the aging of five activated carbons in the containment exhaust air of a German pressurized water reactor to find out whether longer stay times can be obtained with activated carbons other than that usually employed (207B (KI)) in the Federal Republic of Germany. The aging with respect to the retention of methyl iodide (CH3131I) was smaller with activated carbons impregnated with KIsub(x) only than with those impregnated additionally or exclusively with a tertiary amine (e.g. TEDA). However, because of the better performance without aging, the best retention was found with the TEDA (only) impregnant. (orig.)

  14. 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

    OpenAIRE

    Fábio Renato Manzolli Leite; Ticiana Sidorenko de Oliveira Capote; Angela Cristina Cilense Zuanon

    2005-01-01

    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); Clearfi...

  15. 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

  16. The integrated PWR

    International Nuclear Information System (INIS)

    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.)

  17. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  18. 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

  19. 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.)

  20. 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...

  1. 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.)

  2. 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.

  3. 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.

  4. The PWR programme

    International Nuclear Information System (INIS)

    For fueling the PWR type reactors two types of fuel were developed: the UO2 and mixed oxide fuels. To satisfy the demand of the operators of UO2-fuelled power plants a specific industrial organization has been established by Cogema and Framatome: Framagema supplies the technical expertise and sells the fuel; FBFC (Societe Franco-Belge de Fabrication Combustible) is manufacturing the fuel by using particularly the zirconium components produced by Zircotube and Cezus. By making possible the recycling of the materials recovered from the spent fuel reprocessing the MOX (mixed oxide fuels) technology represents an important venture for the future electronuclear sector. To implement this project Cogema created together with Belgonucleaire (the administrator of the Dessel manufacture plant) the GIE COMMOX, in charge with marketing of this fuel. On the other side Cogema which produces MOX in its facility at Cadarache, is at present building the plant at Melox of a capacity of 120 tonnes/year. After presenting the present situation with UO2 and MOX fuels the paper ends with considerations concerning the future fuels and fuels for future and further future reactors

  5. 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...

  6. 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...

  7. 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…

  8. Primer on Social Economics.

    Science.gov (United States)

    Darcy, Robert L.

    An elaboration of the author's booklet entitled "First Steps Toward Economic Understanding," this primer is designed to help the reader develop a functional understanding of the economic process so that he can make wiser decisions on issues of social policy and on matters affecting his economic well-being. The document is not "economics in one…

  9. 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.)

  10. 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

  11. 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.

  12. Condensate purification in PWR reactors

    International Nuclear Information System (INIS)

    The recommendations made by the VGB task group on 'condensate purification for PWR reactors' 1976 are discussed in detail. Techniques and circuiting possibilities of condensate purification for BBR steam generators (forced circulation) and KWU steam generators (U tube with blow-down) are mentioned. (HP)

  13. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  14. PWR AXIAL BURNUP PROFILE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  15. A Quantum Groups Primer

    Science.gov (United States)

    Majid, Shahn

    2002-05-01

    Here is a self-contained introduction to quantum groups as algebraic objects. Based on the author's lecture notes for the Part III pure mathematics course at Cambridge University, the book is suitable as a primary text for graduate courses in quantum groups or supplementary reading for modern courses in advanced algebra. The material assumes knowledge of basic and linear algebra. Some familiarity with semisimple Lie algebras would also be helpful. The volume is a primer for mathematicians but it will also be useful for mathematical physicists.

  16. 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

  17. 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

  18. Solarium primer. [Monograph

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    Solaria can be used to collect solar energy for space heating and provide an open environment. This primer introduces the basic concepts of an energy-efficient solarium and provides information on the practical aspects of siting, design, and construction. The horticultural use of a solarium as a greenhouse is not dealt with except in explaining the influence this may have on its design and operation. The book shows the characteristics of different types of solaria, their advantages as well as their limitations. It also explains the prerequisite of a successful addition to an existing home. Design and construction details accompany the text. 73 references, 2 tables.

  19. Overview of PWR chemistry options

    Energy Technology Data Exchange (ETDEWEB)

    Nordmann, F.; Stutzmann, A.; Bretelle, J.L. [Electricite de France, Central Labs. (France)

    2002-07-01

    EDF Central Laboratories, in charge of engineering in chemistry, of defining the chemistry specifications and studying the operation feedback and improvement for 58 PWR units, have the opportunity to evaluate many options of operation developed and applied all around the world. Thanks to these international relationships and to the benefit of a large feedback from many units, some general evaluation of the various options is discussed in this paper. (authors)

  20. 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)

  1. 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...

  2. Thermodynamic modelling of PWR coolant

    International Nuclear Information System (INIS)

    Spinel solubilities on PWR primary circuit surfaces vary with temperature, pH and coolant H2 concentration. The available solubility data are discussed for Fe, Ni, Co and Zn oxides, and species are identified where data are very limited or absent. An equilibrium thermodynamic model is described to predict the solubility, and results are described predicting relative Fe and Ni solubility under normal operating conditions and during shutdown/startup. The relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels are also considered. (R.P.)

  3. 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

  4. Optimization of control area ventilation systems for Japanese PWR plants

    International Nuclear Information System (INIS)

    The nuclear power plant has been required to reduce the cost for the purpose of making the low-cost energy since several years ago in Japan. The Heating, Ventilating and Air Conditioning system in the nuclear power plant has been also required to reduce its cost. On the other hand the ventilation system should add the improvable function according to the advanced plant design. In response to these different requirements, the ventilation criteria and the design of the ventilation system have been evaluated and optimized in Japanese PWR Plant design. This paper presents the findings of the authors' study

  5. UPTF experiment, flow phenomena during full-scale loop seal clearing of a PWR

    International Nuclear Information System (INIS)

    To investigate the flow phenomena in the primary system of a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) occurring with a small or intermediate break, experiments were performed at the full-scale upper plenum test facility (UPTF). Within the transient and accident management (TRAM) program integral and separate effect tests were carried out to study loop seal clearing and to provide data for the further improvement of computer codes concerning the reactor safety analysis. This paper describes the UPTF tests that focus on the sequence of loop sealclearance in a four-loop operation for two different cold leg break sizes and the residual water levels, the flow patterns in, and the pressure drops across a single loop seal during the clearing. The UPTF results obtained from a single-loop seal operation are compared with experimental data and correlations available in the literature. Two correlations are proposed which allow the quantification of residual water levels in the loop seal under PWR conditions. It is shown that the steam-water test results gained from the full-scale UPTF with realistic PWR loop seal geometry differ from those obtained from the full or small-scale test facilities under air-water conditions. The UPTF experiments indicate the substantial need for steam-water test data from a full-scale facility with realistic PWR geometries in order to validate PWR LOCA thermal-hydraulic system codes to predict loop seal clearing correctly. (orig.)

  6. PWR fuel: experience and development

    International Nuclear Information System (INIS)

    The start-up of the large French nuclear program has rapidly led FRAGEMA to be one of the first PWR fuel suppliers. FRAGEMA is a joint subsidiary of two companies whose scopes of supply are fully complementary: FRAMATOME (NSSS vendor) and COGEMA (nuclear fuel cycle service supplier). At the center of these two activities FRAGEMA is in charge of designing and marketing fuel assemblies. Assistance is also offered to nuclear power plant operators in all fuel related fields by providing a wide range of services and a number of specialized components. Over the past years a statistical data base has been accumulated on fuel assembly behaviour under various operating conditions. At the same time extensive experimental programs have been, set up to develop advanced products to cope with utilities needs in the future. An overview of these two sides of our experience is presented in the following

  7. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  8. 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 ...

  9. 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.)

  10. 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…

  11. 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...

  12. A Primer on Multilevel Modeling

    Science.gov (United States)

    Hayes, Andrew F.

    2006-01-01

    Multilevel modeling (MLM) is growing in use throughout the social sciences. Although daunting from a mathematical perspective, MLM is relatively easy to employ once some basic concepts are understood. In this article, I present a primer on MLM, describing some of these principles and applying them to the analysis of a multilevel data set on…

  13. 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…

  14. 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)

  15. 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…

  16. Maintenance robot for PWR plant

    International Nuclear Information System (INIS)

    The remote operation, automatic machines utilized in the field of the maintenance of component machinery and equipment in nuclear power plants, so-called maintenance robots, have produced effects in the reduction of radiation exposure, the improvement of the quality of working, the shortening of working time and so on, but still many robots have their specialized functions. The expectation of present day society to robots has been diversified, and the technical development of high function robots is advanced positively. In this report, the recent examples of the high function robots developed for PWR power stations with the support of technical progress and the trend of the technical development are explained. The needs and seeds of maintenance robot development are discussed. As the examples of heightening the functions of maintenance robots, the next generation ultrasonic testing machine highly advanced by sensor technology and size and weight reduction mechanism technology, the intelligent monitoring system for welding using AI technology and other manpower-saving robots are shown. (K.I.)

  17. Thermodynamic modelling of PWR coolant

    International Nuclear Information System (INIS)

    Corrosion products released from PWR and VVER primary circuit surface oxides are transported in the coolant to the core, where they deposit and are activated to form radioactive corrosion products, which can be re-released to re-deposit on out-of-core surfaces. Spinel solubilities vary with the pH, temperature and sometimes the hydrogen concentration of the coolant. This paper describes the development of an equilibrium thermodynamic model to predict such changes, and discusses the extent of the available solubility data for Fe, Ni, Co and Zn oxides. Results are described on the relative solubility of Fe and Ni under both normal operating conditions and during shutdown/start-up, and on the relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels. Comparison of the calculated corrosion product concentrations with reactor measurements indicates that, in reactors with low Ni content in the steam generator alloys, the concentration of Ni in the coolant is limited by its availability in the surface oxide. In reactors with high-Ni alloys, the circulating Ni concentrations may be dominated by colloidal material. The calculated changes in Ni and Fe concentrations during the acid-reducing phase of shutdown are in reasonable agreement with measurements from Sizewell B. The paper highlights the need for a more comprehensive open corrosion product data base, the need to consider both boiling and radiolysis in the core on corrosion product solubility in different parts of the primary circuit and, finally, the importance of kinetic factors at low temperature behaviour during shutdown and start-up. (author)

  18. 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.)

  19. 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)

  20. Basic study on PWR plant behavior under the condition of severe accident

    International Nuclear Information System (INIS)

    In this paper, we report on the core cooling effect by natural circulation cooling of the primary cooling system in all core cooling function loss accidents caused by SBO in PWR plant compared with BWR. We also report on the core cooling effect by using air as the final heat sink in place of the seawater by opening the main steam valve of the steam generator. On the other hand, we discuss the behavior of PWR plant in the most serious case that the damage such as LOCA is caused by earthquake and that SBO due to the subsequent tsunami causes the reactor isolation and all function of reactor core cooling system loss. That is the case that LOCA and SBO occur in superimposed manner. We can show the results from the simulation experiments that, in PWR plant, even if it is fell into the reactor core cooling function loss due to SBO, natural circulation cooling can keep the reactor core cool down as long as the feed water is supplied to SG by the turbine-driven auxiliary feed-water pump and also that the cooling effect of even more is expected by ensuring the heat-pass to the atmosphere by opening the main steam valve. We also clarify the plant behaviors under the condition that LOCA and SBO occur in superimposed manner in PWR through the simulation experiments. (author)

  1. 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.

  2. Behaviour of organic iodides under pwr accident conditions

    International Nuclear Information System (INIS)

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH3sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

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

    International Nuclear Information System (INIS)

    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

  4. 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

  5. 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)

  6. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL)

  7. Full MOX core design for PWR

    International Nuclear Information System (INIS)

    Full MOX core design for APWR was analyzed in nuclear design, fuel integrity analysis, thermal hydraulic design and safety analysis et. al. Feasibility of Full MOX core was confirmed from these analyses without any large modifications. Full MOX PWR core has very good characteristics in which single Pu content in an assembly, burnable poison free, higher burnup and longer cycle operation are feasible. (author)

  8. 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.)

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. Sample Return Primer and Handbook

    Science.gov (United States)

    Barrow, Kirk; Cheuvront, Allan; Faris, Grant; Hirst, Edward; Mainland, Nora; McGee, Michael; Szalai, Christine; Vellinga, Joseph; Wahl, Thomas; Williams, Kenneth; Lee, Gentry; Duxbury, Thomas

    2007-01-01

    This three-part Sample Return Primer and Handbook provides a road map for conducting the terminal phase of a sample return mission. The main chapters describe element-by-element analyses and trade studies, as well as required operations plans, procedures, contingencies, interfaces, and corresponding documentation. Based on the experiences of the lead Stardust engineers, the topics include systems engineering (in particular range safety compliance), mission design and navigation, spacecraft hardware and entry, descent, and landing certification, flight and recovery operations, mission assurance and system safety, test and training, and the very important interactions with external support organizations (non-NASA tracking assets, landing site support, and science curation).

  14. 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.

  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. Loop Quantum Geometry: A primer

    OpenAIRE

    Corichi, Alejandro

    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-expert...

  17. 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.

  18. 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.

  19. REWET, PWR LOCA accident experiments

    International Nuclear Information System (INIS)

    1 - Description of test facility: The REWET-II facility was designed for the investigation of the reflooding phase of a LOCA. The main design principle is the accurate simulation of the rod bundle geometry and the primary system elevations. This is necessary in order to have the correct flow channels and hydrostatic pressures for the reflooding process. The reactor vessel is simulated by a stainless steel U-tube construction consisting of downcomer, lower plenum, core and upper plenum. The primary loops contain a pipe simulating the broken loop and a connection line between the upper plenum and the downcomer simulating five intact loops. The containment is simulated by a pressure vessel (not in scale). Steam generators and primary pumps are simulated with flow resistances. The ECC-water can be injected to the downcomer and/or to the upper plenum by a pump or from an accumulator. All the elevation in the reactor vessel simulator are scaled to 1:1 (except the reactor upper head). The scale of the volumes and flow areas is 1:2333 referring to the number of the fuel rod simulators in the facility and the fuel rods in the reference reactor. The rod bundle is either in a hexagonal shroud, the inside distance of the opposite walls is 54.3 mm and the wall thickness 2 mm, or in a round shroud, the inner diameter is 66.0 mm and the wall thickness 2 mm. The simulation of the fuel-rod bundle consists of 19 indirectly electrically heated simulator rods. The heating coils are inside stainless steel claddings in magnesium oxide insulation. The heated length, the outer diameter and the lattice pitch of the fuel-rod simulators as well as the number (= 10) and construction of the rod bundle spacers are the same as in the reference reactor. The upper ends of the rods are attached to the upper tie plate. 2 - Description of test: Pressurized water reactors in use in Finland reactors have certain unique features which make them different from most other PWR designs. The 6 horizontal

  20. Two phase flow in a PWR vessel downcomer during a refill phase of LOCA

    International Nuclear Information System (INIS)

    The refill stage of the Hypothetical Loss of Coolant Accident (LOCA), of a Pressurized Water Reactor (PWR), has been the subject of considerable analysis and experimental study even through it is a very unlikely event. The Large Break Loss of Coolant Accident has been studied extensively in PWR systems to assess the effectiveness of the emergency core cooling system to maintain the fuel rods within safe temperature level. A particular phase of the transient, known as the Refill phase may be reached when the emergency core coolant, inject via the cold legs could be prevented from completely entering the core due to opposing flow of steam in the annular downcomer of the PWR vessel. Under certain conditions the upward flowing steam can hold-up the liquid and by-pass it around the downcomer annulus to the break. Experiments were performed at close to atmospheric pressure in a polycarbonate 1/10th - scale model of a typical four loop Westinghouse pressurized Water Reactor. The objective of the study was to identify hydraulic conditions and flow regimes that exist during the refill stage of a LOCA in the PWR downcomer, and to study the effects of various liquid inlet conditions (non-uniform liquid injection) into the downcomer on the flooding characteristics. The tests were performed using air and water at the working fluids to simulate the refill process under thermal equilibrium conditions. The effect of hot leg penetrations (blockages) on the flooding characteristics was also investigated and studied. The results were plotted using the Wallis J* parameter, and the mean annular circumference w (was used as the characteristic dimension). The flow regimes types and their spatial distribution in the PWR vessel downcomer for air-water counter-current flow were mapped, presented and discussed. The controlling mechanisms for the flooding were postulated and discussed. The flow pattern maps associated with various liquid modes were constructed. In addition the corresponding

  1. Algorithms and Software for PCR Primer Design

    OpenAIRE

    Huang, Yu-ting

    2015-01-01

    Polymerase Chain Reaction (PCR) is a widely used technology in molecular bi-ology for DNA amplification. To generate multiple copies of a DNA molecule, a pair ofprimers (two synthesized DNA sequences with a total length of 15-30 bases) are annealedto the boundaries of the targeted DNA molecule. Then, the new replicated DNA fragmentelongates from one primer to the other.Though primers always hybridize to their respective complements within DNAsequences, primer pairs for targeted DNA sequences ...

  2. 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.

  3. Horizontal Drop of 21- PWR Waste Package

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design

  4. Progress of PWR reactor fuels: OSIRIS equipments

    International Nuclear Information System (INIS)

    The experimental reactor Osiris situated at the Saclay Nuclear Centre is a reactor fitted with tests and monitoring facilities. Of the pool and open core type, it can test the test fuel of PWR power stations under high neutron flux. The characteristic stresses of the operating states of power reactors can be reproduced in experimental devices suited to the various study subjects, be this the creep and deformation of zircaloy claddings, the behavior of fuel rods to power ramps, to load following, to remote regulation, to the cooling state in double phase or just analytical tests. The experimental irradiation devices extend from the single static coolant capsule, such as the NaK alloy, to the dynamic coolant test loop that operates in the cooling conditions representative of PWR's including water chemistry. Ancillary devices make it possible to carry out examinations and non-destructive testing: immersed neutron radiography, gamma scanning visualization monitoring device, eddy currents, profilometering

  5. Thorium fuel cycle study for PWR applications

    International Nuclear Information System (INIS)

    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 UO2 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 ThO2-PuO2 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 UO2 fuel. (author). 6 refs., 3 tabs., 6 figs

  6. Horizontal Drop of 21- PWR Waste Package

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 11) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design

  7. 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.

  8. URPD: a specific product primer design tool

    Directory of Open Access Journals (Sweden)

    Chuang Li-Yeh

    2012-06-01

    Full Text Available Abstract Background Polymerase chain reaction (PCR plays an important role in molecular biology. Primer design fundamentally determines its results. Here, we present a currently available software that is not located in analyzing large sequence but used for a rather straight-forward way of visualizing the primer design process for infrequent users. Findings URPD (yoUR Primer Design, a web-based specific product primer design tool, combines the NCBI Reference Sequences (RefSeq, UCSC In-Silico PCR, memetic algorithm (MA and genetic algorithm (GA primer design methods to obtain specific primer sets. A friendly user interface is accomplished by built-in parameter settings. The incorporated smooth pipeline operations effectively guide both occasional and advanced users. URPD contains an automated process, which produces feasible primer pairs that satisfy the specific needs of the experimental design with practical PCR amplifications. Visual virtual gel electrophoresis and in silico PCR provide a simulated PCR environment. The comparison of Practical gel electrophoresis comparison to virtual gel electrophoresis facilitates and verifies the PCR experiment. Wet-laboratory validation proved that the system provides feasible primers. Conclusions URPD is a user-friendly tool that provides specific primer design results. The pipeline design path makes it easy to operate for beginners. URPD also provides a high throughput primer design function. Moreover, the advanced parameter settings assist sophisticated researchers in performing experiential PCR. Several novel functions, such as a nucleotide accession number template sequence input, local and global specificity estimation, primer pair redesign, user-interactive sequence scale selection, and virtual and practical PCR gel electrophoresis discrepancies have been developed and integrated into URPD. The URPD program is implemented in JAVA and freely available at http://bio.kuas.edu.tw/urpd/.

  9. URPD: a specific product primer design tool

    Science.gov (United States)

    2012-01-01

    Background Polymerase chain reaction (PCR) plays an important role in molecular biology. Primer design fundamentally determines its results. Here, we present a currently available software that is not located in analyzing large sequence but used for a rather straight-forward way of visualizing the primer design process for infrequent users. Findings URPD (yoUR Primer Design), a web-based specific product primer design tool, combines the NCBI Reference Sequences (RefSeq), UCSC In-Silico PCR, memetic algorithm (MA) and genetic algorithm (GA) primer design methods to obtain specific primer sets. A friendly user interface is accomplished by built-in parameter settings. The incorporated smooth pipeline operations effectively guide both occasional and advanced users. URPD contains an automated process, which produces feasible primer pairs that satisfy the specific needs of the experimental design with practical PCR amplifications. Visual virtual gel electrophoresis and in silico PCR provide a simulated PCR environment. The comparison of Practical gel electrophoresis comparison to virtual gel electrophoresis facilitates and verifies the PCR experiment. Wet-laboratory validation proved that the system provides feasible primers. Conclusions URPD is a user-friendly tool that provides specific primer design results. The pipeline design path makes it easy to operate for beginners. URPD also provides a high throughput primer design function. Moreover, the advanced parameter settings assist sophisticated researchers in performing experiential PCR. Several novel functions, such as a nucleotide accession number template sequence input, local and global specificity estimation, primer pair redesign, user-interactive sequence scale selection, and virtual and practical PCR gel electrophoresis discrepancies have been developed and integrated into URPD. The URPD program is implemented in JAVA and freely available at http://bio.kuas.edu.tw/urpd/. PMID:22713312

  10. 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

  11. 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

  12. 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. .

  13. Space-dependent dynamics of PWR

    International Nuclear Information System (INIS)

    The azimuthal dependent reactor dynamics coupled to thermohydraulics are studied by using the neutron-flux and coolant temperature signals measured at an actual PWR. The second azimuthal mode of neutron-flux fluctuation was found, and the coupling of the mode to thermohydraulics of the coolant was suggested. The coherent coolant flow in the reactor core seems to sustain this spatial oscillation mode. (authors)

  14. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  15. 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

  16. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.)

  17. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    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.)

  18. Steam shut-off valves for PWR type reactors

    International Nuclear Information System (INIS)

    Fast acting closure means are requested in PWR type reactors as well as in BWR to safely shut-off the live steam at the turbine input in the event of accident. The design and control system of steam shut-off valves acted by the fluid system and intended for PWR type reactors, are described. The role of these valves in a PWR is discussed with the specified requirements involved

  19. 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)

  20. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  1. Dismantling and decommissioning experience of commercial PWR

    International Nuclear Information System (INIS)

    Regarding the relatively youthness of FRAMATOME PWR's in operation none of these reactor needs to be decommissioned before 1992. However feasibility studies have been carried out by FRAMATOME for an on site entombment of active components and heavy equipments. In the past, partial dismantling of the reactor internals of the CHOOZ reactor: PWR of 320 MWe and a complete removal of the thermal shield protecting the reactor vessel were conducted successfully. After repair, the reactor power output has been upgraded of 10% and the reactor operates satisfactorily since 1970. More recently the discovery of scarce defects affecting centering pins of control guide tube located in the upper reactor internals of 900 MWe plants has initiated the construction of several ''Hot stand equipments'' for the systematic replacement of these centering pins. FRAMATOME is presently actively studying possible options consisting either to extend the plant life beyond its initial licence life, or to convert classical PWR into an advanced reactor more economical in terms of uranium consumption

  2. Pu-breeding feasibility in PWR

    International Nuclear Information System (INIS)

    This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO2 or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived

  3. 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

  4. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  5. 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.

  6. 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.

  7. Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

    International Nuclear Information System (INIS)

    Highlights: • Accidental loss of pool coolant due to leakage in a PWR SFP was studied using MAAP5. • The effect of emergency ventilation on the accident progression was investigated. • The effect of emergency injection on the accident progression was discussed. - Abstract: A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent

  8. 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.

  9. 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.)

  10. PWR Core 2 Project accident analysis

    International Nuclear Information System (INIS)

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  11. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Although 14C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO2), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14C in reactor coolant. A simple chemical kinetic model predicts that CH3OH would be the initial product from radiolytic reactions of 14C following its formation from 17O. CH3OH is predicted to arise as a result of reactions of OH. with CH4 and CH3, and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH3OH can be thermally reduced to CH4 in PWR conditions, although formation of CO2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH4 is the dominant form in PWR and CO2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble material or suspended

  12. 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

  13. 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

  14. 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)

  15. 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.

  16. 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

    are individually analysed in terms of GC content, presence of GC clamp at 3'-end, the risk of primer dimer formation, the risk of intra-primer complementarity (secondary structures) and the presence of polyN stretches. Furthermore, PHUSER offers the option to insert linkers between DNA fragments, as well as highly......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......, 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...

  17. 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)

  18. 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-...

  19. 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.

  20. 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.

  1. 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

  2. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  3. Navy lifts veil on PWR research

    International Nuclear Information System (INIS)

    The author describes the experience of Rolls Royce in developing nuclear reactors for the Navy. Reference is made to the commissioning of HMS Sceptre in February 1978, Britain's 14th nuclear submarine. This event coincided with a decision to lift the veil somewhat on a Research and Development programme that has remained secret for nearly 20 years. Factors that have inhibited progress in this field are mentioned. One of these factors has been the high cost of marine nuclear propulsion systems, tending to limit interest to very large vessels or some special purpose craft. Another factor has been slowness to develop universally acceptable safety criteria, to allow for free and ready access of nuclear vessels to ports. A third factor has been the military origins of much of the development work. A new factor that has arisen recently is the development of the Westinghouse PWR (pressurised water reactor) for marine use in the UK. This has involved collaboration with the US Westinghouse Electric Corporation. Rolls Royce and Associates were chosen to manage this work, which is here described, including the first PWR to be designed and built in Britain and incorporated into a submarine (HMS Vulcan). Much of the design work has been concerned with development of the reactor core and increasing the endurance of the vessel between refuellings. Another aspect was less noise and vibration. Costs of this work are stated, and new test facilities are described. (U.K.)

  4. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  5. PWR and WWER thorium cycle calculation

    International Nuclear Information System (INIS)

    The first step of the investigation of the thorium fuel cycle with HELIOS 1.8 is validation of the results obtained from the code for this particular type of fuel. To complete this first task we performed calculation of the benchmark announced by IAEA in 1995. The benchmark was based on a simplified PWR model of the assembly with reduced fuel composition. This calculation was focused on a comparison of the methods and basic nuclear data. After successful validation of the code we focused our work on calculating the PWR and WWER thorium fuel cycles. The thorium cycle begins after the first use of UO2 fuel in the reactor as separation of plutonium from the burnt fuel. Separated plutonium is mixed with thorium and used as a new nuclear fuel in the reactor. For our calculation we prepared two variants of the assembly - the first variant is a homogeneous distribution and the second one is a non-homogenised distribution of thorium fuel in the assembly. The model of non-homogenised distribution of Pu-Th fuel was designed by replacing selected rods of the classical UO2 assembly by Pu-Th rods. These selected rods are distributed symmetrically in the assembly. Other rods in the assembly remain the same as in the classical UO2 assembly. The calculated and compared values are criticality and fuel composition as a function of burnup (Authors)

  6. Design of Passive Containment Cooling System of PWR using Multi-pod Heat Pipe

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jun Seok; Kim, Sang Nyung [Kyunghee University, Yongin (Korea, Republic of)

    2012-05-15

    As we have seen in the Fukushima nuclear crisis, major accident is not just a hypothetical accident. As in a BWR nuclear power plant, a containment is the last bastion against radiation leakage for a PWR nuclear power plant. The wholesomeness the containment is directly connected with the safety of a nuclear power plant. Therefore, a passive cooling system for the containment is required to prevent a major accident of the nuclear power plant. This study perfected the conceptual design about the Passive Containment Cooling System (including a suppression pool) using Multi-pod Heat Pipe(MPHP) that meets the requirements of air-tightness and economic feasibility

  7. Canadian municipal carbon trading primer

    International Nuclear Information System (INIS)

    The trading of greenhouse gas (GHG) emissions is being suggested as an effective economic way to meet Canada's Kyoto target. Emissions trading is a market-based instrument that can help achieve environmental improvements while using the market to absorb the economical and effective measures to achieve emissions reductions. Placing a value on emissions means that in order to minimize costs, companies will be motivated to apply the lowest-cost emission reductions possible for regulatory approval. The two main types of emissions trading that exist in Canada are the trading of emissions that lead to the formation of smog or acid rain, and the trading of greenhouse gas emissions that lead to climate change. Since carbon dioxide is the most prevalent GHG, making up approximately 75 per cent of Canadian GHG emissions, the trading of units of GHGs is often referred to as carbon trading. The impact that emissions trading will have on municipal operations was the focus of this primer. The trading of GHG involves buying and selling of allowances of GHGs between contracting parties, usually between one party that is short of GHG credits and another that has excess credits. The 3 common approaches to emissions trading include allowance trading (cap and trade), credit trading (baseline and credit), and a hybrid system which combines both credit and allowance trading systems. The issues that impact municipalities include the debate regarding who owns the credits from landfills, particularly if power is generated using landfill gas and the power is sold as green power. Other viable questions were also addressed, including who can claim emission reduction credits if a city implements energy efficiency projects, or fuel substitution programs. Also, will municipalities be allowed to trade internationally, for example, with municipalities in the United States, and how should they spend their money earned from selling credits. This report also presents highlights from 3 emissions

  8. 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)

  9. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  10. Beam shaping for laser initiated optical primers

    Science.gov (United States)

    Lizotte, Todd E.

    2008-08-01

    Remington was one of the first firearm manufacturing companies to file a patent for laser initiated firearms, in 1969. Nearly 40 years later, the development of laser initiated firearms has not become a mainstream technology in the civilian market. Requiring a battery is definitely a short coming, so it is easy to see how such a concept would be problematic. Having a firearm operate reliably and the delivery of laser energy in an efficient manner to ignite the shock-sensitive explosive primer mixtures is a tall task indeed. There has been considerable research on optical element based methods of transferring or compressing laser energy to ignite primer charges, including windows, laser chip primers and various lens shaped windows to focus the laser energy. The focusing of laser light needs to achieve igniting temperatures upwards of >400°C. Many of the patent filings covering this type of technology discuss simple approaches where a single point of light might be sufficient to perform this task. Alternatively a multi-point method might provide better performance, especially for mission critical applications, such as precision military firearms. This paper covers initial design and performance test of the laser beam shaping optics to create simultaneous multiple point ignition locations and a circumferential intense ring for igniting primer charge compounds. A simple initial test of the ring beam shaping technique was evaluated on a standard large caliber primer to determine its effectiveness on igniting the primer material. Several tests were conducted to gauge the feasibility of laser beam shaping, including optic fabrication and mounting on a cartridge, optic durability and functional ignition performance. Initial data will be presented, including testing of optically elements and empirical primer ignition / burn analysis.

  11. 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.)

  12. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235U/UO2 : Pu/ThO2 : 233U/ThO2 - and the conventional recycle-mode uranium system - 235U/UO2 : Pu/UO2. The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  13. 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)

  14. 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)

  15. 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)

  16. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author)

  17. 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

  18. 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)

  19. 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)

  20. Subcooled decompression analysis in PWR LOCA

    International Nuclear Information System (INIS)

    The thermo-hydraulic behavior of the coolant in the primary system of a nuclear reactor is important in the core heat transfer analysis during a hypothetical loss-of-coolant accident (LOCA). The heat transfer correlations are strongly dependent on local thermo-hydraulic conditions of the coolant. The present work allows to calculate such thermo-hydraulic behavior of the coolant during subcooled decompression in PWR LOCA by solving the mass, momentum, and energy conservation equations by the method of characteristics. Detailed studies were made on the transient coolant outflow at the pipe rupture and the effect of frictional loss and heat addition to the coolant on the decompression. Based on the studies, a digital computer code, DEPCO-MULTI, has been prepared and numerical results are compared with the ROSA (JAERI) and the LOFT (NRTS) semiscale test data with various coolant pressures, temperatures, pipe break sizes, and complexity of flow geometry. Good agreement is generally obtained

  1. Recriticality risk in PWR spent fuel pools

    International Nuclear Information System (INIS)

    In this paper we investigated the situation in a PWR Spent Fuel Pool (SFP) following a long-term loss of power / loss of cooling accident. In the SFP there is a large amount of water with soluble boron between the fuel assemblies. There may be a problem from the point of view of criticality safety if the water of the SFP starts to boil and evaporate. A thermal-hydraulic analysis was performed using a simplified model of the SFP. The thermal-hydraulic analysis shows that in all cases a chaotic boiling phenomenon develops. This indicates that even if there is an issue of (near-)criticality, it will have a very intermittent nature. The multiplication factor of the SFP was evaluated with a Monte Carlo calculation. The neutronic analysis was performed for several representative cooling situations. In all cases, the system remains (deeply) subcritical. (author)

  2. Exponential experiments on PWR spent fuel assemblies

    International Nuclear Information System (INIS)

    An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for PWR spent fuel assembly. The axial background neutron flux is measured as a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of Poisson regression to the net induced fission neutron counts. It was revealed that the average exponential decay constants for the C15, J14, G23 and J44 assemblies were determined to be 0.130, 0.127, 0.125 and 0.121, respectively. (author)

  3. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  4. Radiation embrittlement of PWR vessel supports

    International Nuclear Information System (INIS)

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

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

    Science.gov (United States)

    Chadwick, Amy E

    2016-06-01

    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

  6. Changes in 900 MW PWR alarm processing policy

    International Nuclear Information System (INIS)

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs

  7. 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.

  8. Analysis and study on nuclear safety of Mitsubishi PWR

    International Nuclear Information System (INIS)

    Theme of safety analysis and study are changing to reflect the needs at the time. This paper introduces the overall aspects of transient and accident analysis performed and presents typical researches related to safety analysis for Mitsubishi PWR. (author)

  9. 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)

  10. 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.

  11. 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 ...

  12. Microsatellite primers for fungus-growing ants

    DEFF Research Database (Denmark)

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

    2002-01-01

    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.......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...

  13. Microsatellite Primers for Fungus-Growing Ants

    DEFF Research Database (Denmark)

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

    2002-01-01

    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.......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...

  14. 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

  15. Overview of US research related to PWR sump clogging

    International Nuclear Information System (INIS)

    In the framework of research of researches related to the PWR sump clogging in Usa, the author presents the history of GSI-191 (assessment of debris accumulation on PWR sump performance), the research to date (technical assessment, regulatory guide and evaluation guidance, model validation), the current and planned tests (chemical effect and calcium silicate tests, latent debris and downstream effect tests, integrated chemical effect tests, EPRI coatings study). (A.L.B.)

  16. Pressure-relieving devices and it's arrangement for PWR

    International Nuclear Information System (INIS)

    There are four types of PWR pressure-relieving devices: direct acting safety valve, pilot-operated pressure relief valve, power-operated pressure relief valve and safety valve with auxiliaries. The principle of operation, characteristics, arrangement of the pressure-relieving devices for PWR recently used at home and abroad, confidence of discharge, experience in service and developing trend of the devices are introduced. The first and second type of the devices are emphasised

  17. 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.

  18. Primer Design Assistant (PDA): a web-based primer design tool

    OpenAIRE

    Chen, S. H.; Lin, C. Y.; Cho, C.S.; Lo, C.Z.; Hsiung, C. A.

    2003-01-01

    Primer Design Assistant (PDA) is a web interface primer design service combined with thermodynamic theory to evaluate the fitness of primers. It runs in a Linux–Apache–MySQL–PHP structure on a PC equipped with dual CPU (Intel Pentium III 1.4 GHz) and 512 Mb of RAM. A succinct user interface of PDA is accomplished by built-in parameters setting. Advanced options on 5′ GC content, 3′ GC content, dimer check and hairpin check are available. The option of covered region constrains the PCR product...

  19. Forest Interpreter's Primer on Fire Management.

    Science.gov (United States)

    Zelker, Thomas M.

    Specifically prepared for the use of Forest Service field-based interpreters of the management, protection, and use of forest and range resources and the associated human, cultural, and natural history found on these lands, this book is the second in a series of six primers on the multiple use of forest and range resources. Following an…

  20. Scrimer: designing primers from transcriptome data

    Czech Academy of Sciences Publication Activity Database

    Mořkovský, Libor; Pačes, Jan; Rídl, Jakub; Reifová, R.

    2015-01-01

    Roč. 15, č. 6 (2015), s. 1415-1420. ISSN 1755-098X R&D Projects: GA MŠk EE2.3.20.0303 Institutional support: RVO:68081766 ; RVO:68378050 Keywords : next-generation sequencing * primer design * SNaPshot * SNP genotyping * transcriptome Subject RIV: EB - Genetics ; Molecular Biology Impact factor: 3.712, year: 2014

  1. 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)

  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. Loop-mediated amplification accelerated by stem primers.

    Science.gov (United States)

    Gandelman, Olga; Jackson, Rebecca; Kiddle, Guy; Tisi, Laurence

    2011-01-01

    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. PMID:22272122

  4. Criticality calculations with MCNP trademark: A primer

    International Nuclear Information System (INIS)

    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

  5. Rust transformation/rust compatible primers

    Science.gov (United States)

    Emeric, Dario A.; Miller, Christopher E.

    1993-01-01

    Proper surface preparation has been the key to obtain good performance by a surface coating. The major obstacle in preparing a corroded or rusted surface is the complete removal of the contaminants and the corrosion products. Sandblasting has been traditionally used to remove the corrosion products before painting. However, sandblasting can be expensive, may be prohibited by local health regulations and is not applicable in every situation. To get around these obstacles, Industry developed rust converters/rust transformers and rust compatible primers (high solids epoxies). The potential use of these products for military equipment led personnel of the Belvoir Research, Development and Engineering Center (BRDEC) to evaluate the commercially available rust transformers and rust compatible primers. Prior laboratory experience with commercially available rust converters, as well as field studies in Hawaii and Puerto Rico, revealed poor performance, several inherent limitations, and lack of reliability. It was obvious from our studies that the performance of rust converting products was more dependent on the amount and type of rust present, as well as the degree of permeability of the coating, than on the product's ability to form an organometallic complex with the rust. Based on these results, it was decided that the Military should develop their own rust converter formulation and specification. The compound described in the specification is for use on a rusted surface before the application of an organic coating (bituminous compounds, primer or topcoat). These coatings should end the need for sandblasting or the removing of the adherent corrosion products. They also will prepare the surface for the application of the organic coating. Several commercially available rust compatible primers (RCP) were also tested using corroded surfaces. All of the evaluated RCP failed our laboratory tests for primers.

  6. 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

  7. 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

  8. Westinghouse advanced passive 600 MWe PWR design

    International Nuclear Information System (INIS)

    Although there has been a sharp downturn in the ordering of commercial nuclear power plants throughout the world, it is nonetheless anticipated that this form of energy will remain vital to the economy of many nations in a long term. One of the important new development activities is that of small plants incorporating passive safety features. The small plants have the merits in terms of low total capital requirement and potentially short lead time. The Electric Power Research Institute sponsored the development of an advanced LWR plant in a nine month Westinghouse program, which terminated in March, 1986. Further development at Westinghouse is now in progress on this design called AP 600 under the sponsorship of the U.S. Department of Energy. On the basis of the proven 600 MWe PWR plant design, the specific design improvement for increased safety and operational margin, reduced plant capital and operating cost, simplified plant systems and components, and increased certainty of meeting construction schedule and cost is pursued. The Westinghouse two-loop plants are very competitive, and the operating performance is outstanding by the comparison of plant capacity factor. The operation and maintenance costs are low. The specific design and the features of modification and improvement are discussed. (Kako, I)

  9. 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

  10. Computer aided information system for a PWR

    International Nuclear Information System (INIS)

    The computer aided information system (CAIS) is designed with a view to improve the performance of the operator. CAIS assists the plant operator in an advisory and support role, thereby reducing the workload level and potential human errors. The CAIS as explained here has been designed for a PWR type KLT- 40 used in Floating Nuclear Power Stations (FNPS). However the underlying philosophy evolved in designing the CAIS can be suitably adopted for other type of nuclear power plants too (BWR, PHWR). Operator information is divided into three broad categories: a) continuously available information b) automatically available information and c) on demand information. Two in number touch screens are provided on the main control panel. One is earmarked for continuously available information and the other is dedicated for automatically available information. Both the screens can be used at the operator's discretion for on-demand information. Automatically available information screen overrides the on-demand information screens. In addition to the above, CAIS has the features of event sequence recording, disturbance recording and information documentation. CAIS design ensures that the operator is not overburdened with excess and unnecessary information, but at the same time adequate and well formatted information is available. (author). 5 refs., 4 figs

  11. Aging effects in PWR power plants components

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  12. Enriched Gadolinium as burnable absorber for PWR

    International Nuclear Information System (INIS)

    This paper is a summary of a master of thesis work in reactor physics made by Ola Seveborn. The work was done at Vattenfall Braensle AB and Ola was guided through the work by the corresponding author of this paper. The results presented are calculations for Ringhals 3, which is a Westinghouse 3-loop PWR within the Vattenfall Group. The fuel is characterized by 17x17 assemblies of AFA type containing 3.80-3.95 w/o 235U and 8 rods containing 2 w/o Gadolinium with an enrichment of 70 w/o 157Gd. The calculations were performed with the Studsvik-Scandpower code package based on the CASMO-4 lattice code and the SIMULATE-3 nodal code. The results are compared to the corresponding calculations for fuel with 5 w/o gadolinium with natural isotopic constitution. The depletion of the cores was done separately for the reference and enriched case. The results show that the gains in average for the five cycles studied are about 70 EFPH per cycle. This is an effect of the lower gadolinium content needed. Also less parasitic absorption of enriched gadolinium in the end of the fuel life contributes to the increased cycle lengths. The abruptly increased reactivity and internal power peaking factor around 10 MWd/kgU do not affect the core design negatively. (authors)

  13. 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.)

  14. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    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

  15. PWR core monitoring system and benchmarking

    International Nuclear Information System (INIS)

    The PWR Power Shape Monitoring System (PSMS) provides site engineers with new capabilities for monitoring and predicting core power distributions. These capabilities can lead to increased plant output as a result of greater operating margins, better load maneuvering, earlier detection of anomalies, and improved fuel reliability. The heart of the PSMS consists of nodal code (NODEP-2/THERM-P) that computes the 3-D core power distribution. This code is coupled to a simplified nodal version of the COBRA-IIIC/MIT-2 thermal-hydraulic model to determine the DNBR. These calculations can be completed in about 30 seconds on a PRIME-750 mini computer. Activation of the calculations and review of the results is through user-friendly interactive software that can be tailored to the requirements and capabilities of the different categories of users through table-driven menus. The PSMS provides unique advances over core power monitoring systems based purely on measurements. The PSMS approach permits the three-dimensional core simulation model to be routinely corrected with in-core/ex-core measurements while simultaneously identifying consistent instrument errors

  16. Tritium management in PWR fuel reprocessing plants

    International Nuclear Information System (INIS)

    Activity, quantity and nature of tritium compounds obtained during head end process (cutting and dissolution) are determined to estimate environmental release hazards in fuel reprocessing plants. Measurements on representative PWR reactor fuels (burnup 33,000 MWdt-1, specific power 30 MW dt-1) show that about 60% of the tritium produced in the reactor diffuses in the cladding where it is fixed. Remaining tritium stays in the irradiated oxide and is found as tritiated water in the solution obtained during fuel dissolution. In the UP3 plant at La Hague (France) tritiated water is disposed into the sea without environmental problems. In the case of a reprocessing plant far from the sea, the PUREX process is slightly modified for concentration of tritium in a limited amount of water (TRILEX process). It is verified experimentally in αβγ lab on actual fuel and by simulation at the pilot seale that the supplementary step ''tritium washing'' of the solvent can be obtained in pulsed columns. 4 tables, 7 figs

  17. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  18. Scaling Analysis for PWR Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuquan [State Nuclear Power Technology R and D Center, Beijing (China); Ye, Zishen [Tsinghua University, Beijing (China)

    2011-08-15

    To test the nuclear power plant safety system performance and verify the relative safety analysis code, a widely used approach is to design and construct a scaled model based on a scaling methodology. For a pressurized water reactor (PWR), the SG scaling analysis is important before designing a scale model which is expected to well simulate the system response to the accident. In this work, a review of the transient process in SG during a loss of coolant accident (LOCA) is first presented, and then a brief natural circulation scaling analysis is performed to get the basic SG scaling design rules. The U-tube scaling design shows the scaling will enlarge the thermal center height ratio while keeping the length ratio when the scale model uses a different diameter ratio and the height ratio, which causes distortion in natural circulation simulation. And then, by the heat transfer scaling analysis, a relation between the U-tube diameter ratio and model height ratio is obtained, and it shows the diameter ratio decreases with the decreasing model height ratio. In the end, the SG transition from the heat sink to the heat source is analyzed, and the results show the SG secondary inventory and the total material heat capacity need to be properly scaled to represent the transition correctly.

  19. Scaling Analysis for PWR Steam Generator

    International Nuclear Information System (INIS)

    To test the nuclear power plant safety system performance and verify the relative safety analysis code, a widely used approach is to design and construct a scaled model based on a scaling methodology. For a pressurized water reactor (PWR), the SG scaling analysis is important before designing a scale model which is expected to well simulate the system response to the accident. In this work, a review of the transient process in SG during a loss of coolant accident (LOCA) is first presented, and then a brief natural circulation scaling analysis is performed to get the basic SG scaling design rules. The U-tube scaling design shows the scaling will enlarge the thermal center height ratio while keeping the length ratio when the scale model uses a different diameter ratio and the height ratio, which causes distortion in natural circulation simulation. And then, by the heat transfer scaling analysis, a relation between the U-tube diameter ratio and model height ratio is obtained, and it shows the diameter ratio decreases with the decreasing model height ratio. In the end, the SG transition from the heat sink to the heat source is analyzed, and the results show the SG secondary inventory and the total material heat capacity need to be properly scaled to represent the transition correctly

  20. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  1. 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.

  2. Aging effects in PWR power plants components

    International Nuclear Information System (INIS)

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

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

    OpenAIRE

    Wang Xiaolei; Hang Xingyi; Li Zhifeng; Wu Yonghong; Lu Yiming; Wang Wen; Qu Wubin; Shen Zhiyong; Zhao Dongsheng; Zhang Chenggang

    2010-01-01

    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 nons...

  4. 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)

  5. 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)

  6. Seismic qualification of PWR plant auxiliary feedwater systems

    Energy Technology Data Exchange (ETDEWEB)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    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.

  7. 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

  8. 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

  9. The development of flow test technology for PWR fuel assemblies

    International Nuclear Information System (INIS)

    The objective of this project is to design and construct a high temperature and pressure flow test facility and to develop flow test technology for the evaluation of PWR fuel performance. For the nuclear fuel safety aspect it is of importance to evaluate the thermalhydraulic compatibility and mechanical integrity of a newly designed fuel through the design verification test. The PWR-Hot Test Loop facility is under construction to be used to perform a pressure drop test, a lift force test and a fretting corrosion test of a fullsize PWR fuel assembly at reactor operating conditions. This facility was designed to be used to produce the hydraulic parameters of the existing PWR fuel assemblies(14x14FA, 16x16FA, 17x17FA) and to verify a design of advanced fuel assemblies (KAFA-I and KAFA-II) developed by KAERI. The PWR-Cold Test Loop facility with the 5x5 Rod Bundles in the test section was designed and installed to carry out the flow distribution study by means of Laser Doppler Velocimeter. The LDV techniques have been developed and used to measure the flow velocity and turbulent intensity for evaluating mixing effects of a newly designed spacer grid with and without mixing vanes, cross flow between the fuel assemblies and a turbulent model. (Author)

  10. 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.

  11. 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

  12. 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.

  13. A primer for criticality calculations with DANTSYS

    Energy Technology Data Exchange (ETDEWEB)

    Busch, R.D. [Univ. of New Mexico, Albuquerque, NM (United States). Nuclear Criticality Safety Group

    1997-08-01

    With the closure of many experimental facilities, the nuclear safety analyst has to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. Although deterministic methods often do not provide exact models of a system, a substantial amount of reliable information on nuclear systems can be obtained using these 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 (UNM) in cooperation with the Radiation Transport Group at Los Alamos National Laboratory (LANL) has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. DANTSYS is the new name of the group of codes formerly known as: ONEDANT, TWODANT, TWOHEX, TWOGQ, and THREEDANT. The primer is designed to teach bu example, with each example illustrating two or three DANTSYS features useful in criticality analyses. Starting with a Quickstart chapter, the primer gives an overview of the basic requirements for DANTSYS input and allows the user to quickly run a simple criticality problem with DANTSYS. Each chapter has a list of basic objectives at the beginning identifying the goal of the chapter and the individual DANTSYS features covered in detail in the chapter example problems. On completion of the primer, it is expected that the user will be comfortable doing criticality calculations with DANTSYS and can handle 60--80% of the situations that normally arise in a facility. The primary provides a set of input files that can be selective modified by the user to fit each particular problem.

  14. A primer for criticality calculations with DANTSYS

    International Nuclear Information System (INIS)

    With the closure of many experimental facilities, the nuclear safety analyst has to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. Although deterministic methods often do not provide exact models of a system, a substantial amount of reliable information on nuclear systems can be obtained using these 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 (UNM) in cooperation with the Radiation Transport Group at Los Alamos National Laboratory (LANL) has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. DANTSYS is the new name of the group of codes formerly known as: ONEDANT, TWODANT, TWOHEX, TWOGQ, and THREEDANT. The primer is designed to teach bu example, with each example illustrating two or three DANTSYS features useful in criticality analyses. Starting with a Quickstart chapter, the primer gives an overview of the basic requirements for DANTSYS input and allows the user to quickly run a simple criticality problem with DANTSYS. Each chapter has a list of basic objectives at the beginning identifying the goal of the chapter and the individual DANTSYS features covered in detail in the chapter example problems. On completion of the primer, it is expected that the user will be comfortable doing criticality calculations with DANTSYS and can handle 60--80% of the situations that normally arise in a facility. The primary provides a set of input files that can be selective modified by the user to fit each particular problem

  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. 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...

  17. 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...

  18. Long-Term Corrosion Behavior of CVD SiC in PWR-Simulating Water

    International Nuclear Information System (INIS)

    SiC ceramics show an outstanding oxidation resistance and a low hydrogen liberation rate in hot steam compared with the current Zr alloys, which promise larger safety margins under severe accident conditions. Moreover, its high temperature strength and stability under high neutron doses also provide high burn-up capability. In spite of the potential benefits of the SiC composite cladding, there are a lot of technical issues that need to be clarified for the LWR application. Especially, the corrosion resistance in the PWR water is an important parameter to insure the cladding performance under normal operating condition. Generally, SiC ceramics are highly corrosion resistant by forming a protective SiO2 layer in an air atmosphere. Long-term corrosion behavior of CVD SiC was investigated under simulated PWR primary water conditions. The dissolved hydrogen of about 2.7 ppm dramatically reduced the corrosion rate of SiC at 360 .deg. C under 20 MPa. The corrosion weight loss of CVD SiC under the dissolved hydrogen-controlled condition was extremely small compared with the condition without controlling the dissolved hydrogen. The oxide layer on the SiC specimen was readily dissolved and the SiC bare surface was exposed at the initial stage of the corrosion test. After the dissolution of the oxide layer, however, further corrosion occurred in a very sluggish way

  19. 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)

  20. Multidimensional two-phase flow regime distribution in a PWR downcomer during an LBLOCA refill phase

    International Nuclear Information System (INIS)

    The multidimensional countercurrent two-phase flow regimes that occur in a pressurized-water reactor (PWR) vessel downcomer during the refill phase of a large-break loss-of-coolant accident are studied using a transparent 1/10 scale model of a PWR vessel. The various flow regimes and their distribution in the downcomer have been identified and mapped for a range of air-water flooding experiments. The two-phase flow patterns that are identified in the downcomer included various types of film flows, droplet flows, countercurrent churn flows and cocurrent flows depending on the flooding condition. Through observation of the two-phase flow dynamics it was deduced that the physical mechanisms associated with the flooding processes could be separated into a liquid entrainment process and a film flow reversal process. In addition to the above exercise, the effect of non-uniform injection of water into the downcomer via different combinations of cold leg was studied similarly by determining flooding curves and flow pattern maps. It was found that differences in the flooding characteristic were noticeable for various water inlet configurations when compared with the uniform injection case. The differences could be explained qualitatively in terms of the flooding mechanisms identified previously by examining the flow patterns in the downcomer for the non-uniform injection tests. ((orig.))

  1. 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.

  2. 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

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

    International Nuclear Information System (INIS)

    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

  4. 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.

  5. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  6. Affecting factors analysis of major equipment erection key path in PWR NPP

    International Nuclear Information System (INIS)

    The affecting factors of major equipment erection in PWR NPP exist impersonally, especially the design and equipment supply has produced some effects on major equipment erection of PWR nuclear power plant. Through the analysis of key path and affecting factors on major equipment erection of PWR NPP, the paper puts forward some countermeasures. (authors)

  7. 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

  8. 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.)

  9. PWR fuel performance and future trend in Japan

    International Nuclear Information System (INIS)

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of RandD on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important RandD targets are the burnup extension, Gd contained fuel, utilization and the load follow capability

  10. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  11. Timing analysis of PWR fuel pin failures

    International Nuclear Information System (INIS)

    This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PF1/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design burnup. Using peaking factors commensurate with actual burnups would result in longer intervals for both reactor designs. This document provides appendices K and L of this report which provide plots for the timing analysis of PWR fuel pin failures for Oconee and Seabrook respectively

  12. 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)

  13. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    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).

  14. 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)

  15. 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

  16. 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.

  17. Evolution of reactor monitoring and protection systems for PWR

    International Nuclear Information System (INIS)

    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.)

  18. Corrosion product transfer in PWR primary circuits during cold shutdowns

    International Nuclear Information System (INIS)

    Two experimental tests have been performed to study the corrosion product transfer during PWR cold shutdowns: one with nickel ferrite and the other one with metallic nickel. The temperature evolution together with boron and oxygen concentration evolutions are similar to those obtained during cold shutdowns of PWR primary circuits. With metallic nickel, the increase of the Ni concentration occurs during the decrease of the primary temperature and mainly when the oxygenation is realised. Whereas, with nickel ferrite, the Ni concentration increase occurs during the 24 hours after the oxygenation. These results compared to the plant data lead to conclude that metallic nickel presence in the core is the most probable hypothesis. (author)

  19. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

  20. 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

  1. 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)

  2. Examination of dissimilar metal welds in BWR and PWR piping

    International Nuclear Information System (INIS)

    This paper addresses dissimilar metal weld examinations at PWRS. Surveys were conducted to document the dissimilar metal weld configurations at PWR plants and to update the information known about dissimilar metal weld configurations at BWR plants. The experiences which BWR utilities have had with dissimilar metal weld examinations are documented and include: correct identification of IGSCC, indications thought to be IGSCC but were actually fabrication flaws, and difficulties encountered with the examination of dissimilar metal welds after stress improvement. An experimental program was conducted which verified that the longitudinal wave procedures developed for BWRs are also applicable to PWR designs

  3. 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

  4. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE

  5. Approximation for maximum pressure calculation in containment of PWR reactors

    International Nuclear Information System (INIS)

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author)

  6. Post DNB heat transfer experiments for PWR fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis. (author)

  7. 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.

  8. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  9. 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.

  10. 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

  11. Methodology for the LABIHS PWR simulator modernization

    International Nuclear Information System (INIS)

    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

  12. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 2800C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 2800C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  13. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel. Pt. 2

    International Nuclear Information System (INIS)

    In previous experiments, done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel, the instrument tubes support structure built into the vault was not included. It consists of a number of grids made up of fairly massive steel girders. These have now been added to the model and experiments performed using water to simulate molten core debris assumed to have fallen on to the vault floor and high-pressure air to simulate the discharge of steam or gas from the assumed breach at the bottom of the pressure vessel. The results show that the tubes support structure considerably reduces the carry-over of liquid via the vault access shafts. (author)

  14. 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.

  15. 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,

  16. 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

  17. Primer izpeljave analize besedila v kvalitativni raziskavi

    OpenAIRE

    Roblek, Vasja

    2013-01-01

    V članku na podlagi kvalitativne raziskave prikazujemo primer načina analize in razlage besedila. V prvem delu se osredotočimo na teoretično opredelitev analitičnega orodja ter značilnosti analize in interpretacije besedil znotraj kvalitativne raziskave. V nadaljevanju na podlagi izsledkov (polstrukturiranih intervjujev in osebnih zapisov, opazovanja delovanja dveh mrež) prikažemo možnost analize in interpretacije dobljenih podatkov z uporabo analitičnega orodja tematske mreže.

  18. Primer izpeljave analize besedila v kvalitativni raziskavi:

    OpenAIRE

    Roblek, Vasja

    2009-01-01

    V članku na podlagi kvalitativne raziskave prikazujemo primer načina analize in razlage besedila. V prvem delu se osredotočimo na teoretično opredelitev analitičnega orodja ter značilnosti analize in interpretacije besedil znotraj kvalitativne raziskave. V nadaljevanju na podlagi izsledkov (polstrukturiranih intervjujev in osebnih zapisov, opazovanja delovanja dveh mrež) prikažemo možnost analize in interpretacije dobljenih podatkov z uporabo analitičnega orodja tematske mreže.

  19. 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

  20. Hygrometric measurement for on-line monitoring of PWR vessel head penetrations

    International Nuclear Information System (INIS)

    In September 1991, a small leak was found on one of the reactor's upper vessel head penetrations. After inspection, other non-throughwall cracks were localized in the lower part of the vessel head adapter in questions. The same type of crack was later found inside some adapters on other French PWR units. After repairs, the safety authorities granted approval to continue unit operation, with the specific provision that a system for ongoing monitoring of the penetrations be set up. Two types of system were selected to detect leaks through any potential cracks: the first is based on nitrogen-13 detection and the second on steam detection. Both systems call for sampling the air in a confined space above the vessel head. The number and distribution of sampling taps in the circuit, and the balancing of their respective flow rates, are factors in proper monitoring of all vessel head penetrations. Gas-injection holes are also installed in the confined space. These holes are used during the sampling system qualification tests to simulate leaks in various positions and calculate the effective performance of the sampling system. Leaks are simulated using a helium-base gas tracer and measuring tracer concentrations in the sampling system. The system for measuring steam levels in air samples uses chilled-mirror hygrometers. A microcomputer takes regular readings, drives the various automatic functions of the measurement system and automatically analyses the readings so as to monitor operations and trigger an alarm at the first sign of a leak. This system has now been installed for a year and a half on three French PWR units and is functioning satisfactorily. (authors). 5 figs

  1. Simulations of PWR spray systems by ASTEC computer code

    International Nuclear Information System (INIS)

    In this paper, sequence of loss of feedwater of steam generators on a PWR 900 MWe is performed by application of integral code ASTEC. The influence of the spray on evolution and source term of severe accident in the containment and in the environment is mainly studied. The results are helpful for the investigation of mitigative measures of severe accident. (authors)

  2. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    International Nuclear Information System (INIS)

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  3. PWR power plant reactor maintenance: site experience and technology transfer

    International Nuclear Information System (INIS)

    PWR reactor maintenance activities has considerably expanded during the last few years at Framatome. The services offered by Framatome can be divided into three main categories: - Implementation of backfits aimed at performance and safety improvement and equipment reliability - Technical assistance for plant operators, especially during refuelling - Maintenance and repair services during both scheduled and unscheduled outages

  4. PWR auxiliary systems, safety and emergency systems, accident analysis, operation

    International Nuclear Information System (INIS)

    The author presents a description of PWR auxiliary systems like volume control, boric acid control, coolant purification, -degassing, -storage and -treatment system and waste processing systems. Residual heat removal systems, emergency systems and containment designs are discussed. As an accident analysis the author gives a survey over malfunctions and disturbances in the field of reactor operations. (TK)

  5. 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

  6. Sub Channel Thermal hydraulics Design Analysis of PWR-KSNP

    International Nuclear Information System (INIS)

    Sub channel analysis for the fuel element of thermal hydraulics design PWR-KSNP reactor has been carried out. PWR-KSNP reactor is a kind of Pressurized Water Reactor (PWR) Nuclear Power Plant developed by Korea (Korean Standard Nuclear Plant), that produce an electricity power about 1000 MWe. In the analysis, a fuel assembly with 4 fuel rods piled up into matrix 2 x 2, and surrounding by 9 sub channels of coolant, was used as a calculation model. There are 3 models of fuel assembly, i.e. the radial factors in the first model are 1.144, 1.144, 1.120 and 1.121, in the second fuel model are 0.994 , 1.005 , 0.987 and 0.989, and in the third model are 2.500, 1.144, 1.120 and 1.121, respectively. The calculated results using the COBRA IV-I code showed that the maximum cladding temperature revolved by 340.3 - 349.0 ℃, the maximum temperature of meat surface (outer of meat) revolved by 498.1 - 758.2 ℃ and the maximum temperature of meat center revolved by 928.5 - 1843.7 ℃, respectively. Whereas the safety margin against DNBR revolved by 6.50 - 2.05. By maximum meat temperature limit of 2804 ℃ and the minimum DNBR of 1.30, it is concluded that the PWR-KSNP design was in the range of safety. (author)

  7. Latest technologies of design and construction for new PWR plants

    International Nuclear Information System (INIS)

    Mitsubishi Heavy Industries, Ltd. (MHI) has so far constructed 23 PWR plants including Mihama Unit-1 Unit, which started operation in 1970. After Genkai Unit-4, which started operation in 1997, construction of PWR plants temporarily halted in Japan. However, the safety assessment of Tomari Unit-3 (3 loops, output 912 MWe) of Hokkaido Electric Power Company started in Nov. 2000, and a safety assessment of Tsuruga Units-3 and 4 (4 loops, output 1538 MWe/unit) of Japan Atomic Power Company, the first APWR units, are also planned to start very soon. PWR plants have now been designed full scale and construction of new PWR plants is going to commence in earnest in Japan. A brief introduction of the new design is followed by a description of the improved design method, upgrade of the 3D-CAD design system and improved project control procedure. The concept of the new construction method planned for the new plants is also introduced. Through these, and based on construction experience of 23 plants, MHI will introduce enhancements to the reliability and safety of the new plants and promote their smooth design and construction

  8. 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)

  9. 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)

  10. Contribution to study and design of PWR plant simulation code

    International Nuclear Information System (INIS)

    This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants

  11. 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

  12. 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.)

  13. Thermohydraulic and constructional boundary conditions of an advanced PWR reactor

    International Nuclear Information System (INIS)

    The advantages and special features of an advanced PWR reactor (FDWR) have been systematically investigated for several years by the Department of Space Flight and Reactor Technology of the University of Brunswick (LRR-TUBS). The FDWR will have a homogeneous core, i.e. the fuel elements will consist of fuel rods of the same size and enrichment. (orig./GL)

  14. 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

  15. Process and device for residual heat removal in a PWR

    International Nuclear Information System (INIS)

    This new process for residual heat remover in a PWR is characterized by the use of the pressurizer relief tank. After cooling by the steam generator, the primary water is taken at the upper part of the pressurizer, then expanded and condensed in the pressurized relief tank. A pump recycles this water and introduces it in the primary circuit

  16. The inertisation of PWR containments by injection of liquid carbondioxide

    Energy Technology Data Exchange (ETDEWEB)

    Karwat, H.; Stolze, P. [Technical University Munich, Garching (Germany)

    1997-03-01

    the transient distribution until uniform concentration of the atmospheric components steam, air and carbondioxide is reached. Time intervals necessary to achieve the required inerting status of approx. 60 vol% may easily be below 30 minutes. The combination of inertisation with the installation of a number of catalytic recombiners remains as a supplementary option. A limited number of experimental results has been made available by fire fighting industry to facilitate the application of the modified code for validation purposes. Comparisons between analyses and experimental results are satisfactory. However, some tests, more typical for PWR containments would be desireable to confirm the post-accident inertisation simulations. (author)

  17. The inertisation of PWR containments by injection of liquid carbondioxide

    International Nuclear Information System (INIS)

    distribution until uniform concentration of the atmospheric components steam, air and carbondioxide is reached. Time intervals necessary to achieve the required inerting status of approx. 60 vol% may easily be below 30 minutes. The combination of inertisation with the installation of a number of catalytic recombiners remains as a supplementary option. A limited number of experimental results has been made available by fire fighting industry to facilitate the application of the modified code for validation purposes. Comparisons between analyses and experimental results are satisfactory. However, some tests, more typical for PWR containments would be desireable to confirm the post-accident inertisation simulations. (author)

  18. Safety criteria and their comparison between WWER and PWR

    International Nuclear Information System (INIS)

    Relevant PWR and WWER fuel safety related criteria are identified and classified into: (a) safety criteria, (b) operational criteria, and (c) design criteria. All criteria and their basis were reviewed in details. PWR and WWER fuel safety criteria are found in principle compatible (very similar if not identical basis) - thus, no fundamental distinction in the approach to defining fuel safety criteria between Eastern and Western Europe appears to exist. Some differences have been observed between the individual criteria and/or their numerical values. These are mostly due to differences in fuel design and materials. The variety in reactor characteristics and country specific licensing requirements sometimes lead to differences between criteria definitions. The report captures the characteristics of PWR/WWER fuel safety and serves as a basis to outline the general safety case for PWR and WWER fuel. The report also serves to highlight the basic fuel safety principles and their bases, and broadens the insight in failure. Especially in view of high burnup and new materials, there is a need to further develop the safety criteria and their numerical values. Even without such innovations, there is a clear potential for refining and improving some of the fuel safety related criteria and/or their numerical values. This report could serve as a basis for discussions on a more in-depth co-operation for future activities needed to verify the existing safety criteria or to support improvements. Closer collaboration in the review of PWR (LWR) and WWER safety criteria and in reviewing the respective safety cases is recommended

  19. 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)

  20. 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.

  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. 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...

  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. PWR-GALE, Radioactive Gaseous Release and Liquid Release from PWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: The PWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from pressurized water reactors (PWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

  5. Coverage evaluation of universal bacterial primers using the metagenomic datasets

    Directory of Open Access Journals (Sweden)

    Mao Dan-Ping

    2012-05-01

    Full Text Available Abstract Background The coverage of universal primers for the bacterial 16S rRNA gene plays a crucial role in the correct understanding of microbial community structure. However, existing studies on primer coverage are limited by the lack of appropriate databases and are restricted to the domain level. Additionally, most studies do not account for the positional effect of single primer-template mismatches. In this study, we used 7 metagenomic datasets as well as the Ribosomal Database Project (RDP to assess the coverage of 8 widely used bacterial primers. Results The coverage rates for bacterial primers were found to be overestimated by previous studies that only investigated the RDP because of PCR amplification bias in the sequence composition of the dataset. In the RDP, the non-coverage rates for all primers except 27F were ≪6%, while in the metagenomic datasets, most were ≫10%. If one considers that a single mismatch near the 3′ end of the primer might greatly reduce PCR efficiency, then some phylum non-coverage rates would change by more than 20%. Primer binding-site sequence variants that could not pair with their corresponding primers are discussed. Conclusions Our study revealed the potential bias introduced by the use of universal bacterial primers in the assessment of microbial communities. With the development of high-throughput, next-generation sequencing techniques, it will become feasible to sequence more of the hypervariable regions of the bacterial 16S rRNA gene. This, in turn, will lead to the more frequent use of the primers discussed here.

  6. 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.

  7. 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...

  8. Complex and adaptive dynamical systems a primer

    CERN Document Server

    Gros, Claudius

    2007-01-01

    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...

  9. 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...

  10. 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 ...

  11. 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

  12. 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.}

  13. MCNPTM criticality primer and training experiences

    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, the analyst may have little experience with the specific codes available at his or her facility. Usually, the codes are quite complex, black boxes capable of analyzing numerous problems with a myriad of input options. Documentation for these codes is designed to cover all the possible configurations and types of analyses but does not give much detail on any particular type of analysis. For criticality calculations, the user of a code is primarily interested in the value of the effective multiplication factor for a system (keff). Most codes will provide this, and truckloads of other information that may be less pertinent to criticality calculations. Based on discussions with code users in the nuclear criticality safety community, it was decided that a simple document discussing the ins and outs of criticality calculations with specific codes would be quite useful. The Transport Methods Group, XTM, at Los Alamos National Laboratory (LANL) decided to develop a primer for criticality calculations with their Monte Carlo code, MCNP. This was a joint task between LANL with a knowledge and understanding of the nuances and capabilities of MCNP and the University of New Mexico with a knowledge and understanding of nuclear criticality safety calculations and educating first time users of neutronics calculations. The initial problem was that the MCNP manual just contained too much information. Almost everything one needs to know about MCNP can be found in the manual; the problem is that there is more information than a user requires to do a simple keff calculation. The basic concept of the primer was to distill the manual to create a document whose only focus was criticality calculations using MCNP

  14. Identification of ryanodine receptor isoforms in prostate DU-145, LNCaP, and PWR-1E cells.

    Science.gov (United States)

    Kobylewski, Sarah E; Henderson, Kimberly A; Eckhert, Curtis D

    2012-08-24

    The ryanodine receptor (RyR) is a large, intracellular calcium (Ca(2+)) channel that is associated with several accessory proteins and is an important component of a cell's ability to respond to changes in the environment. Three isoforms of the RyR exist and are well documented for skeletal and cardiac muscle and the brain, but the isoforms in non-excitable cells are poorly understood. The aggressiveness of breast cancers in women has been positively correlated with the expression of the RyR in breast tumor tissue, but it is unknown if this is limited to specific isoforms. Identification and characterization of RyRs in cancer models is important in understanding the role of the RyR channel complex in cancer and as a potential therapeutic target. The objective of this report was to identify the RyR isoforms expressed in widely used prostate cancer cell lines, DU-145 and LNCaP, and the non-tumorigenic prostate cell line, PWR-1E. Oligonucleotide primers specific for each isoform were used in semi-quantitative and real-time PCR to determine the identification and expression levels of the RyR isoforms. RyR1 was expressed in the highest amount in DU-145 tumor cells, expression was 0.48-fold in the non-tumor cell line PWR-1E compared to DU-145 cells, and no expression was observed in LNCaP tumor cells. DU-145 cells had the lowest expression of RyR2. The expression was 26- and 15-fold higher in LNCaP and PWR-1E cells, respectively. RyR3 expression was not observed in any of the cell lines. All cell types released Ca(2+) in response to caffeine showing they had functional RyRs. Total cellular RyR-associated Ca(2+) release is determined by both the number of activated RyRs and its accessory proteins which modulate the receptor. Our results suggest that the correlation between the expression of the RyR and tumor aggression is not related to specific RyR isoforms, but may be related to the activity and number of receptors. PMID:22846571

  15. PWR accident management realated tests: some Bethsy results

    International Nuclear Information System (INIS)

    The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

  16. 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

  17. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U3 O8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  18. 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

  19. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  20. Training of the PWR staff at EdF

    International Nuclear Information System (INIS)

    Electricite de France (EdF) is now operating thirty-two 900-MW and nine 1300-MW pressurized water reactor (PWR) units and a 1500-MW surgenerator; two 900-MW and ten 1300-MW PWR units are under construction. The number of persons employed to run these plants is at present 24000. A very important training program provided initial knowledge, specific training, and retraining in order to ensure nuclear safety. The following list includes those features required in the most viable and efficient training program: first contacts for familiarization with company and safety training (including radiation protection in nuclear plants); industrial adaptation, depending on the specialty; training for the function performed; skill maintenance and refresher courses in the function; and preparatory or promotion courses when performing a new function

  1. The modeling of flooding conditions in a PWR downcomer

    International Nuclear Information System (INIS)

    In this paper the modeling capability of the two fluid model is investigated with respect to the flooding phenomena that occur during the refill phase in a PWR experiencing a large cold leg break loss of coolant accident. A review of the literature indicate's that the basic modeling requirements have not been met and that it is still not apparent that a two fluid approach is acceptable. A modeling study is presented that attempts to investigate the applicability of a two dimensional two fluid model to the two phase hydrodynamic conditions that occur in the downcomer. A specific flow map has been constructed from experimental flooding tests on a 1/10 scale PWR and has been used to apply closure relationships appropriate for the type of flow regimes experienced during the experimental investigations. The model was validated by comparison with experimental test data and found to give good results when compared with experimental flooding curves

  2. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  3. 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)

  4. 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)

  5. PWRDYN: a computer code for PWR plant dynamic analysis

    International Nuclear Information System (INIS)

    This report describes analytical models and calculated results of a PWR plant dynamic analysis code PWRDYN. The code has been developed in order to analyze and evaluate transient responses for small disturbance such as operating mode change and control system characteristic analysis. The features included in PWRDYN are 1) One loop approximation of primary loops, 2) Praimary coolant is always subcooled, 3) At the secondary side of steam generator is used one dimensional model and natural circulation is calculated assuming constant by positive driving head. 4) Main control systems are incorporated. In the transient responses caused by small perturbation, the calculated results by PWRDYN are in good agreement with the RETRAN calculations. Furthermore, computing time is very short so as about one seventh of real time, hence the code is convenient and useful for dynamic analysis of PWR plants. (author)

  6. Bond graph model for prediction of PWR natural circulation

    International Nuclear Information System (INIS)

    In operation of a Pressurized Water Reactor, natural circulation is an efficient, passive heat transfer mechanism for cooling. It is often employed for heat removal in operational transients, especially in long-term decay-heat removal operation. To simulate the dynamics of natural circulation, a bond graph representation of the PWR primary system and causal manipulation of the field equations has been modeled. The bond graph method calls for establishing the dynamic equations of multiport systems in state-variable form. Using the analogy of circuit elements in electrical networks, a bond graph consists of multiport capacitances, inertances, dissipations, sources of effort and flow, transformers, gyrators, and ideal junction elements. By treating each component in a PWR primary loop as a multiport element in the bond graph, a set of state-space equations representing the thermal/hydraulic responses of the loop is obtained. The state equations are then solved iteratively by using the program DYSIS developed by MIT

  7. 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

  8. Steam explosion simplified modeling. Pressure assessment in a PWR cavity

    International Nuclear Information System (INIS)

    A simplified approach of steam explosion modeling in a PWR cavity has been developed in the framework of an R and D program supported by the EDF Basic Design Division (SEPTEN). It can be used as a stand-alone engineering tool (called MGEV) or included in an integrated code (MAAP for example). This approach is mainly based on an expansion model with compressibility and inertial effects. Heat and mass transfers at the high pressure bubble interface are assessed by the EXCOBUL model and the corium mass in the interaction zone is estimated by the Park-Corradini model. The implementation aims at the calculation of pressure records on the cavity boundaries. These results will be used for mechanical evaluations which should demonstrate that the potential damages in a large PWR would be limited to the cavity region

  9. 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)

  10. 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)

  11. 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

  12. The development of 1530 MW steam turbine for Advanced PWR

    International Nuclear Information System (INIS)

    MITSUBISHI has been manufacturing 27 nuclear steam turbines and total output is over 20,000 MW since the 1970 first delivery of nuclear steam turbine. Based on these our successful experiences, MITSUBISHI is making a continuous effort to develop the most modern steam turbine with the lager capacity, higher efficiency and higher reliability. And now, the first Advanced PWR is being planned to be built at Tsuruga No.3 and No.4 by Japan Atomic Power Co. as the largest plant with an electric power of about 1530 MW. To apply this Advanced PWR plants, we are going forward planing and developing the largest capacity nuclear steam turbine. This paper shows the key technologies of target capacity nuclear steam turbine such as 54 inches low pressure last blade, and the advanced technologies to realize high performance and high reliability steam turbine. (author)

  13. Polynomial parameterized representation of macroscopic cross section for PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fiel, Joao Claudio B., E-mail: fiel@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and {sup 235} U {sub 92} enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K{sub inf}, generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

  14. Assessment of subcriticality during PWR-type reactor refueling

    International Nuclear Information System (INIS)

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-α and Feynman-α methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  15. 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

  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. 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)

  18. Training of PWR operators in Electricite de France

    International Nuclear Information System (INIS)

    Electricite de France has started up in 1981 eight PWR units (900 MWe) and five or six more will be commissioned in the course of each of the next few years. Such a programme will require an unprecedented effort of recruitment and training by the company. New staff will be recruited partly by taking on young school-leavers and partly by internal recruitment within EDF of staff working in conventional thermal or nuclear power stations. Training will be provided each year to around 4,000 staff. The paper reviews successively the following points: organizing the operation of a power station with four PWR units (900 MWe); the criteria for personnel selection; training programmes suited to the origin and function of operating personnel; training staff to carry out a given function; the existing training resources

  19. Calculation of drop course of control rod assembly in PWR

    International Nuclear Information System (INIS)

    The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

  20. Dependence of Corrosion Behavior of SiC Ceramics on PWR Water Chemistry

    International Nuclear Information System (INIS)

    There have also been efforts on applying the SiCf/SiC composites to the light water reactor (LWR) fuel cladding and guide tubes as well as channel boxes for fuel assembly of the boiling water reactor (BWR). In spite of potential benefits of the SiC composite cladding, there are a lot of technical issues that need to be clarified for the LWR application because the previous research on the SiCf/SiC composite has mostly been focused on high-temperature application. Especially, the corrosion resistance in the PWR water is an important parameter to insure the cladding performance under normal operating condition. Generally, SiC ceramics are highly corrosion resistant by forming a protective SiO2 layer in an air atmosphere. However, the corrosion resistance of SiC is largely dependent on the fabrication route of SiC and the exact environmental condition. For example, in a high-temperature and high-pressure water or a steam environment, the corrosion resistance of SiC is decreased because the protectiveness of the SiO2 layer is deteriorated. Until now, the exact corrosion behavior of SiC in high temperature water is not clear due to the differences in the test samples and conditions. The kinetics of the SiC corrosion in an LWR condition, therefore, needs to be defined to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In this study, we evaluated the corrosion behavior of SiC-based ceramics fabricated by various routes. We also examined the effect of PWR water chemistry on the corrosion resistance of SiC

  1. Low temperature crack propagation of nickel-based weld metals in hydrogenated PWR primary water

    International Nuclear Information System (INIS)

    The effect of hydrogenated PWR primary water on the Low Temperature Crack Propagation (LTCP) susceptibility of nickel-based weld metals Alloy 182, 82, 152 and 52 was studied performing J-R tests at a slow displacement rate in simulated low temperature PWR primary water. When tested in an environment with high hydrogen content (100 cm3 H2/kg H2O), all the studied materials showed a remarkable decrease in the fracture toughness (JIC or JQ) values compared with the air test results. Alloy 182 showed the lowest average fracture toughness values in each test environment. The results obtained at a lower hydrogen content (∼ 30 cm3 H2/kg H2O) suggest, that Alloy 182 is the most susceptible nickel-based weld metal to LTCP, especially at low hydrogen contents. Intergranular cracking was predominant when the JIC value was low. Test results of pure weld metal Alloys 182 and 52 were also compared with the results of dissimilar metal weld (DMW) specimens of Alloy 182 and 52. The pure weld metals were substantially more susceptible to LTCP than the DMW specimens. Pre-exposure to high temperature hydrogenated water did not affect the fracture toughness of any of the test materials. The degradation in toughness is assumed to be caused by a hydrogen-induced intergranular cracking mechanism, where the precipitates in the weld metals acting as hydrogen trapping sites play an important role. The LTCP susceptibility of the studied alloys is discussed based on the present hydrogen embrittlement mechanisms. (authors)

  2. 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.

  3. 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.

  4. 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

  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. 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.)

  7. Fire experiences: principal lessons learned, application in PWR power plants

    International Nuclear Information System (INIS)

    The article reviews the principal design rules to be borne in mind for PWR nuclear units installation. These rule takes into account: the specific character of materials involved (safety aspect for nuclear construction), experience acquired as a result of fires in EDF production units, and the results obtained from tests carried out by the EDF at Fort de Chelles between 1980 and 1982, especially in the field of PVC cables

  8. DNB experiments for high-conversion PWR core design

    International Nuclear Information System (INIS)

    It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion pressurized water reactors (PWR). To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid Freon 12 and water under the actual operating conditions. In addition, DNB heat flux measurements in an annular-flow channel were carried out for the design of the fertile rods, which are installed in thimble tubes. (orig.)

  9. DNB experiments for high conversion PWR core design

    International Nuclear Information System (INIS)

    It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion PWR design. To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid freon-12 and the actual water. And also DNB heat flux mesurements in an annular flow channel were carried out for design of fertile rods which are installed in thimble tubes. (orig.)

  10. 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

  11. Ciclon: A neutronic fuel management program for PWR's consecutive cycles

    International Nuclear Information System (INIS)

    The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author)

  12. SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

    International Nuclear Information System (INIS)

    1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

  13. Safety philosophy and concepts for future European PWR

    International Nuclear Information System (INIS)

    PWR are presently the most developed nuclear technology in the world and particularly in Europe. The existing units have shown during twenty years of operation a high level of reliability and safety. Despite the remarkable result already achieved the necessity of regaining public confidence leads nuclear industry to look for further safety improvements. The european utilities, convinced of the necessity of future nuclear capacity investments, have decided to unify their efforts for the preparation of the next nuclear generation

  14. 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.

  15. A practical method for optimization of fuel management of PWR

    International Nuclear Information System (INIS)

    A practical method for simulation of fuel management optimization of PWR cores with two-dimensional model is described. The general objective of the optimization is to choose a set of refuelling arrangement schemes, which will produce the maximum economic profit on condition that it meets the safety criteria of PWR. It oftern requires quite a lot of computer time to simulate the optimized schemes. An effective and acceptable optimization strategy, two-step search method, has been developed. The first step of algorithm consists of several approaches based on the information avilable and the past experiences with refuelling. The second step allows a further improvement of the previously determined optimum schemes. The maximum radial power peaking factor, Wp, is defined as the objective function. Several physical criteria are examined to propose the constraints. The main intention is to minimize, the objective function Wp, subjected to various constraints. Hence a computer program, 2DFEOF in FORTRAN 77, was developed. Some calculations were done for a typical PWR core on an IBM-4341 computer. The satisfactory results were obtained at reasonable low computational costs. It spent nearly 9 mins CPU time for 3 fuel cycles with a 1/8 core configuration

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

    Science.gov (United States)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    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.

  17. 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)

  18. 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

  19. Design study of long-life PWR using thorium cycle

    Science.gov (United States)

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-01

    Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that 231Pa better than 237Np as burnable poisons in thorium fuel system. Thorium oxide system with 8% 233U enrichment and 7.6˜ 8% 231Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1% Δk/k and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53% Δk/k and reduced power peaking during its operation.

  20. 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.

  1. 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

  2. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  3. 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.

  4. 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.

  5. Analysis of the estimated isotopic concentration of PWR spent fuel

    International Nuclear Information System (INIS)

    Using SCALE4.4 SAS2H, HELIOS and CASMO codes, isotopic inventories in PWR spent fuel have been calculated and compared with the reported experimental data. 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. Influences of correction factors to the multiplication factor have also been investigated. The calculated biases and uncertainties of U-235 in PWR spent fuel seem to be 2.8 % / 3.9 %, -2.0 % / 4.1 % and 5.0 % / 4.5 %. In the case of transuranium isotopes and fission products, the results calculated by HELIOS and CASMO codes show a large discrepancy from the reported experimental data in comparison of SAS2H results. In general it is believed that SAS2H is better than HELIOS and CASMO for estimating isotopic inventory in PWR spent fuel. It is revealed that correction factors obtained by codes of interest give rise to the maximum difference of about 0.05 in the multiplication factor

  6. 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.

  7. 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.

  8. 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.

  9. Study on virtual simulation technology for operation and control of PWR

    International Nuclear Information System (INIS)

    The way to build graphical models of PWR with MultiGen Creator is discussed, and the three-dimensional model used in the virtual simulation is built. The mathematical simulation model for PWR based on the platform of MFC and Vega is built through the analysis of the mathematical simulation of PWR. The way to perform the virtual effect is introduced associating with the Pressurizer. And, all above parts are connected in one with VC++ to perform the whole virtual simulation of PWR. (authors)

  10. Relativistic Astrophysics and Cosmology: A Primer

    International Nuclear Information System (INIS)

    '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 the text. Their students will be in the author

  11. A primer on climate relevant technology transfer

    International Nuclear Information System (INIS)

    The aim of this primer is to support technology transfer initiatives related to mitigation of emissions of greenhouse gases and adaptation to climate change in developing countries and economies in transition. It is meant to provide guidance to all parties involved in the transfer of climate relevant technologies, micluding government, business organisations and other non-governmental organizations representing all sectors relevant for climate change. The emphasis in this primer is on strategic and organisational issues related to the transfer of climate relevant technologies, in particular with a view to utilise climate-relevant technologies in the development process of developing countries and countries in economic transition. It builds upon insight gained in National Needs Assessments regarding the Transfer of Enviromentally Sound Technologies to developing countries. Chapter 2 deals with the technology transfer process for climate relevant technologies. It provides a framework for co-ordinated action of the stakeholders involved in the technology transfer process, with a view to improve the utilisation of climate relevant technologies in the development process of developing countries. It therefore proposes three main 'pillars' for climate relevant technology transfer, i.e. creating an enabling environment for stakeholders' participation, assessing mitigation and adaptation needs and opportunities, and implementing and evaluating mitigation and adaptation actions. Chapter 3 elaborates on creating an enabling environment for stakeholders' participation. It deals in particular with the selection of priority sectors and obtaining involvement of stakeholders representing these prioritised sectors, in order to start a roundtable process for the preparation of mitigation and adaptation actions. Chapter 4 covers the second pillar 'assessing mitigation and adaptation needs and opportunities'. This encompasses several assessment tasks including: identification and

  12. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

  13. Computer analyses on loop seal clearing experiment at PWR PACTEL

    International Nuclear Information System (INIS)

    Highlights: • Code analyses of loop seal clearing experiment with PWR PACTEL are introduced. • TRACE and APROS system codes are used in the analyses. • Main events of the experiment are well predicted with both codes. • Discrepancies are observed on the secondary side and in the core region. • Loop seal clearing phenomenon is well simulated with both codes. - Abstract: Water seal formation in the loop seal in pressurized water reactors can occur during a small or intermediate break loss-of-coolant accident, causing temporary fuel overheating. Quantification of the accuracy of overheating prediction is of interest in the best-estimate safety analyses, even though the peak cladding temperatures due to the water seal formation in the loop seal seldom approach acceptance criteria as such. The aim of this study was to test and evaluate the accuracy with which the thermal–hydraulic system code nodalizations of the PWR PACTEL predict loop seal clearing in a small break loss-of-coolant-accident test performed with the PWR PACTEL facility. PWR PACTEL is a thermal–hydraulic test facility with two loops and vertical inverted U-tube steam generators. Post-test simulations were performed with the TRACE and APROS system codes. In the post-test simulations, the main events of the transient such as the decrease in the core water level, depressurization of the primary circuit, and the behavior of the water seal formation and clearing in the loop seal were predicted satisfactorily by both codes. However, discrepancies with the experiment results were observed in the analyses with both codes, for example the core temperature excursions were halted too early and the peak temperature predictions were too low. The core water level increase caused by loop seal clearing was overestimated with both codes, and the pressure and temperature were overestimated on the secondary side of the steam generators. Loop Seal 2 was evidently cleared out while Loop Seal 1 remained closed

  14. Void distribution analysis with high-speed x-ray CT scanner for design of PWR grid spacer

    International Nuclear Information System (INIS)

    A grid spacer of PWR fuel plays a dominant role to increase a thermal margin for safety operation of PWR since cooling effects of fuel rod are significantly promoted by fluid mixing induced by mixing vanes of grid spacers. Recently, CFD has become an available tool for the prediction of DNB performance of designed grid spacers by examining a fluid mixing effect. However, for complicated flow path like PWR fuel assembly, the current CFD is not applicable to the prediction of a two-phase slug flow due to the existence of complicated structure of interfaces between gas and liquid phases. As a result of that, an alternative approach is required to compensate the shortage of CFD applicability to the evaluation of DNB performance in the slug flow condition. In the present study, void distributions of air and water two-phase flows inside a 3x3 test rod bundle were experimentally investigated to examine the effects of grid spacers on the slug flow. A high-speed X-ray CT scanner was employed as a measurement tool that can provide with instantaneous void distributions in a cross section. Grid spacers of different types were used and the measured data were examined to investigate the differences of void distributions. Since it can be assumed that the DNB performance will be influenced by void distribution, our water DNB test data were referred to take into account the relationship between the void distributions and the performance of resistance to DNB. As a result, the detailed examinations of void profiles in flow subchannels provide the corresponding void distributions with the results of the water DNB tests. It was confirmed that the void distributions obtained by the high speed X-ray CT scanner could yield one of the available aspects from the view point of flow fields when we compare the DNB performance under two-phase slug flow condition between designed grid spacers. (author)

  15. 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)

  16. 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

  17. 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.

  18. Iodine partition coefficient measurements at simulated PWR steam generator conditions: Interim data report

    International Nuclear Information System (INIS)

    Iodine partition coefficients (defined as the ratio of the concentration of iodine species in the aqueous solution to the iodine concentration in the vapor phase) were measured at simulated PWR steam generator conditions (2850C and 6.9 MPa), using carrier-free radioactive 131I in the form of sodium iodide. The iodine tracer concentration was maintained at ∼6 x 10-11 mol/L; boric acid concentration was varied from 0 to 0.4 mol/L; and the solution pH (measured at 250C) was adjusted from 4 to 9 by the addition of lithium hydroxide. Iodine partition coefficients decrease with increasing boric acid concentration; however, the iodine volatility is essentially independent of the solution pH for a given boric acid concentration. Sparging the solutions with air at room temperature increases the iodine volatility by an order of magnitude, compared to that achieved with argon sparging. Iodine partition coefficient measurements ranged from a low of 200 (in 0.2 M boric acid sparged with air) to 400,000 (in purified water sparged with argon)

  19. 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.

  20. 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.)

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

    International Nuclear Information System (INIS)

    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.)

  2. 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.

  3. The evaluation of erosion-corrosion problems in Taiwan PWR carbon steel piping

    International Nuclear Information System (INIS)

    Taiwan PWR Nuclear Power Plant Units 1 and 2 implemented the projects of Pipe Wall Thinning Measurement under the request of ROCAEC to prevent the events due to the piping erosion/corrosion. The purpose of this paper is to present the improvements in the evaluation method for the identification of the potential piping systems and components in PWR

  4. 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

  5. PRIMER FOR FINANCIAL ANALYSIS OF POLLUTION PREVENTION PROJECTS

    Science.gov (United States)

    This primer will serve as a basic guide to pollution prevention investment -- specifically, the preparation of financial comparisons and justifications for such expenditures. he emphasis is on the basic analytical techniques needed to justify pollution prevention investments. onc...

  6. PREP-PWR-1.0: a WIMS-D/4 pre-processor code for the generation of data for PWR fuel assemblies

    International Nuclear Information System (INIS)

    The PREP-PWR-1.0 computer code is a substantially modified version of the PREWIM code which formed part of the original MARIA System (Report J.E.N. 543). PREP-PWR-1.0 is a comprehensive pre-processor code which generates input data for the WIMS-D/4.1 code (Report PEL 294) for PWR fuel assemblies, with or without control and burnable poison rods. This data is generated at various base and off-base conditions. The overall cross section generation methodology is described, followed by a brief overview of the model. Aspects of the base/off-base calculational scheme are outlined. Additional features of the code are described while the input data format of PREP-PWR-1.0 is listed. The sample problems and suggestions for further improvements to the code are also described. 2 figs., 2 tabs., 12 refs

  7. Integral type small PWR with stand-alone safety

    Energy Technology Data Exchange (ETDEWEB)

    Makihara, Yoshiaki [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2001-09-01

    A feasibility study is achieved on an integral type small PWR with stand-alone safety. It is designed to have the following features. (1) The coolant does not leak out at any accidental condition. (2) The fuel failure does never occur while it is supposed on the large scale PWR at the design base accident. (3) At any accidental condition the safety is secured without any support from the outside (stand-alone safety secure). (4) It has self-regulating characteristics and easy controllability. The above features can be satisfied by integrate the steam generator and CRDM in the reactor vessel while the pipe line break has to be considered on the conventional PWR. Several counter measures are planned to satisfy the above features. The economy feature is also attained by several simplifications such as (1) elimination of main coolant piping and pressurizer by the integration of primary cooling system and self-pressurizing, (2) elimination of RCP by application of natural circulating system, (3) elimination of ECCS and accumulator by application of static safety system, (4) large scale volume reduction of the container vessel by application of integrated primary cooling system, (5) elimination of boric acid treatment by deletion of chemical shim. The long operation period such as 10 years can be attained by the application of Gd fuel in one batch refueling. The construction period can be shortened by the standardizing the design and the introduction of modular component system. Furthermore the applicability of the reduced modulation core is also considered. (K. Tsuchihashi)

  8. 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)

  9. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation

  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. Effect of startup ramp rate on PWR fuel reliability

    International Nuclear Information System (INIS)

    A wide range of startup strategies and restart times currently exists for commercially operated pressurized water reactors (PWRs). The variability in PWR restart strategies is a function of several factors, including reactor system instrument calibration, primary and secondary water chemistry control, and vendor specified fuel rod ramp rate limitations. Fuel vendors, as a means to mitigate pellet-cladding interaction (PCI) leading to fuel rod failures, specify reactor power ramp rate limitations following a refueling outage. Typical restart ramp rates range between 3% per hour and 4% per hour of full reactor power above a threshold reactor power level between 20% and 40% full power. This paper summarizes an analytical evaluation performed to assess the technical basis for PWR restart ramp rate restrictions and to provide the technical justification to propose less restrictive power ramp rate conditions. Two combinations of PWR reactor types (Yonggwang Unit 2 and 4) and fuel rod designs were used to evaluate the impact of ramp rate and threshold power conditions on the PCI behavior of once-burned and twice-burned fuel rods. The fuel rod condition at the reactor restart of interest was established using the ESCORE steady state fuel performance program. Detailed PCI calculations were performed using the FREY fuel rod behavior program. The assessment identified significant margin to PCI failure for current ramp rate conditions used in YGN Unit 2 and 4. Based on the analytical evaluation presented, ramp rates up to 5% per hour above threshold power levels up to 60% of full reactor power can be used without concern for fuel rod integrity during reactor restarts following a refueling outage

  12. Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

    International Nuclear Information System (INIS)

    An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U3O8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U3O8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author)

  13. Loop-Mediated Amplification Accelerated by Stem Primers

    OpenAIRE

    Laurence Tisi; Guy Kiddle; Olga Gandelman; Rebecca Jackson

    2011-01-01

    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 ...

  14. PD5: A General Purpose Library for Primer Design Software

    OpenAIRE

    Riley, Michael C.; Aubrey, Wayne; Young, Michael; Clare, Amanda

    2013-01-01

    Background Complex PCR applications for large genome-scale projects require fast, reliable and often highly sophisticated primer design software applications. Presently, such applications use pipelining methods to utilise many third party applications and this involves file parsing, interfacing and data conversion, which is slow and prone to error. A fully integrated suite of software tools for primer design would considerably improve the development time, the processing speed, and the reliab...

  15. Short-term calculations to supplement the RS 16 B PWR experiments with internals (PWR1 to PWR5), using the LECK 4 computer code

    International Nuclear Information System (INIS)

    Within the framework of research project RS 16 B sponsored by the German BMFT a series of a blowdown experiments, DWR1 to DWR5, were performed using a vessel with dummy internals under conditions similar to those in a PWR. The prime objective of these experiments was the investigation of the highly transient blowdown phenomena in the discharge nozzle and the determination of the induced loads on the internals. As a partner in the project, KWU carried out both pre-test predictions and post-test analyses of these experiments using, among others, the computer code LECK 4. For the most severe blowdown test DWR5, the influence of the most important model parameters on the blowdown analysis was investigated in detail. These investigations suggest that, similar to the long-term analyses, calculations using the homogeneous critical flow model would improve agreement between calculation and experiment. (orig./RW)

  16. Vertical Drop Of 21-PWR Waste Package On Unyielding Surface

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only

  17. Probabilistic consequence analysis of ATWSs in a PWR plant

    International Nuclear Information System (INIS)

    PWR responses (in terms of overpressures, DNBR and other safety-related quantities) to ATWSs are being probabilistically investigated by applying response surface methodology to ALMOD, a computer program for simulation of large amplitude transients. The reactor considered for the analysis is the 1300 MWel reference KWU reactor plant. A comprehensive set of input quantities--including operational, engineering and physical variables--is taken into account. Results are presented for the first phases of station-blackout and loss-of-heat sink ATWSs

  18. 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)

  19. 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.

  20. Application of H∞ control theory to PWR power control

    International Nuclear Information System (INIS)

    In this paper, a robust controller is designed by the use of H∞ control theory for the PWR power control. The design specification is incorporated by the frequency weights using the mixed-sensitivity problem. The robustness of H∞ control is verified by comparing with the classical output feedback control and LQG control in the case of measurement delay of the power measurement system. The H∞ optimal control shows excellent stability-robustness and performance-robustness for external disturbances and noises, model parameter variations, and modeling errors. It also provides a practical design method because the design specification can be easily implemented

  1. 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)

  2. Design build process flow visualization model plant PLTN PWR type

    International Nuclear Information System (INIS)

    Scale-down version of nuclear power plant type PWR model and process flow visualization has been design and constructed. This scale-down model includes primary and secondary cooling systems, and transmission line in three dimensional layout with a 1: 33,33 scale. The construction of scale model has been done in five steps that are study literature, field survey, drawing scale design, construction, and test. The results is scale-down model integrated with monitoring system using lab view and interlock system using PLC. The test result shows that process flow has operated as required in design specification. (author)

  3. Friends of the Earth case against the PWR

    International Nuclear Information System (INIS)

    Friends of the Earth's case against Sizewell B has been summarised in a report entitled 'Critical Decision: Should Britain buy the Pressurised Water Reactor?'. This showed that on economic and safety grounds, Sizewell B would not be a good choice for the electricity consumer or the country at large. Events since the end of the Inquiry, particularly those affecting the economic case, have confirmed this conclusion. This paper will summarise the case, both during the Inquiry and subsequent to this, as well as make reference to the long-term environmental implications of the Central Electricity Generating Board's PWR programme. (author)

  4. 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

  5. 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)

  6. 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.)

  7. 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

  8. Neutronal aspects of PWR control for transient load following

    International Nuclear Information System (INIS)

    The purpose of this thesis is to qualify the CRONOS diffusion code on a load transient in grey mode control. First of all, we have established a general axial calculational model and studied the important physical phenomena: xenon oscillation, grey rods absorption, radial leaks modelling, effect of the initial conditions in Iodine and Xenon. In a second stage, a three dimensional calculation has been performed, the results of which have been compared to a PWR 900 TRICASTIN 3 experiment and have been in good agreement. In the last part, we show that the results of the axial model using one-dimensional CRONOS calculations are quite consistent with the three-dimensional calculation

  9. Quality surveillance for PWR power plant reactor internals manufacturing

    International Nuclear Information System (INIS)

    The structure and the function of the reactor internals of the improved generation Ⅱ PWR power plant is instructed briefly, the critical factors and difficulties in the manufacture process for reactor internals are analyzed, the quality control and surveillance of reactor internals manufacturing is discussed, especially the critical factors and difficulties of the quality control in the manufacture process for the main parts of reactor internals and in the reactor internals assembling process are represented in detail, the key points of the resident manufacture supervision is presented, and other key points of quality control in the manufacture process are also given, such as the documents control and personnel control. (author)

  10. For sale: 7 AGR stations and a brand new PWR

    International Nuclear Information System (INIS)

    Britain's seven AGR stations and the Sizewell B PWR will pass to private ownership under the UK government's plan to privatise the two nuclear generators, Nuclear Electric and Scottish Nuclear, sometime next year. Under the new set-up, the two generators will become operating subsidiaries of a holding company which will be headquartered in Scotland. The companies' ageing Magnox gas-cooled reactors will remain in a separate public sector company before being transferred to British Nuclear Fuels (BNFL) at the time of privatisation. (author)

  11. Concept of safety systems for next generation PWR (APWR+)

    International Nuclear Information System (INIS)

    The concept of the next generation PWR, which is expected to come after the APWR and is named the APWR+, is being studied, considering that the light water reactors are seemed to be dominant also in the 21st century. The APWR+ is designed to have the features of four-train safety systems, divergent emergency electrical sources, and passive core cooling system using steam generators at early stage of the Loss of Coolant Accident. The basic concept has been made, and more detailed investigation is scheduled in near future. (author)

  12. Recent progress in SG level control in French PWR plants

    International Nuclear Information System (INIS)

    Controlling the steam generator (SG) level is of major importance in a large PWR plant. This has led to extensive work on SG computer models. This paper presents results of the comparison between calculations and tests on the first four-loop plant in France. Four-loop plants started up after 1985 will be equipped with digital instead of analog controllers. A new SG level control has been designed and then optimised using the validated SG model. A prototype of this new system has been successfully tested on a three-loop plant. 4 refs

  13. 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

  14. Improvement in PWR flexibility the french program 1975-1995

    International Nuclear Information System (INIS)

    Between 1975 and 1985, a substantial effort was launched in France to greatly improve PWR's flexibility, resulting in the current situation where both frequency control and load follow are now routinely performed on most plants in operation. Based on rapidly accumulating operational experience and on all expertise acquired in the past decade, a second-generation core control strategy is now being finalized for application on all future 1400 MW plants (with commercial operation scheduled in 1992 for first unit). This 20-year program is discussed

  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.; 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)

  16. 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.

  17. 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.

  18. 21-PWR Waste Package Side and End Impacts

    International Nuclear Information System (INIS)

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1

  19. PWR fuel performance and burnup extension programme in Japan

    International Nuclear Information System (INIS)

    Since the first PWR nuclear power plant Mihama Unit 1 initiated commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts on improving the technology of PWRs. The results can already be seen by the significantly improved performance of the PWR plants now in operation. Mitsubishi Heavy Industries, Ltd supplied the nuclear fuel assemblies, which now amount to almost 5000. Although some trouble with fuel was experienced in the beginning, the progressive efforts made to improve the fuel design and manufacturing technology have resulted in the superior performance of Mitsubishi fuels. Since fuel of current design should comply with the limitation set in Japan for a maximum discharged fuel assembly average burnup of less than 39,000 MW·d/t, the maximum burnup is now around 37,000 MW·d/t. However, an increase in this burnup limitation has been strongly requested by Japanese utilities in order to make nuclear power more economic and thus more competitive with other power generation methods. A summary is given of the design improvements made on Mitsubishi fuel, as well as demonstration programmes of current design fuel to prove its superior reliability and to prepare the database for a future extension of burnup. (author)

  20. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    International Nuclear Information System (INIS)

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated

  1. 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.

  2. 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)

  3. Modifications needed to operate PWR's plants in G-Mode

    International Nuclear Information System (INIS)

    The production of electricity from PWR nuclear plants represents 44% of the total production of electricity in France for 1984, and 68% of the electricity produced by Thermal power plants (127 TWh over 187 TWh). These data show clearly that the French PWR plants do not work in ''base mode'' anymore but have to fit production with consumption, in other words to assume the frequency control. To participate permanently to the load follow and frequency control, it appeared that some improvements in the field of pressurizer level and pressure control were necessary as well as in the field of operator aids computer. It should be noted that these improvements are useful even without taking into account the constraints due to load follow and frequency control because of the mechanical stress in the CVCS piping, for instance. Some additional tests are planned to better identify this specific problem. The need of a more flexible operating mode than ones given by the initial system (black control rods), significantly reduced in 1973 due to the application of the ECCS criterion, led EDF and Framatome to develop a new operating mode (G. Mode) allowing a faster power escalation (5% PN/mn) whatever the fuel burn-up. This new operating mode improves significantly also the flexibility of operation when the frequency control is needed, and helps a lot the operators in such cases. All the 900 MWe Nuclear plants will be able to operate in ''G mode'' before the end of 1984

  4. 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)

  5. 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)

  6. DNB analysis with mechanistic models for PWR fuel assemblies

    International Nuclear Information System (INIS)

    In order to predict the DNB heat flux of PWR fuel assemblies and the critical power of BWR fuel bundles, the Boiling Transition Analysis Code CAPE' has been developed in the IMPACT project. The CAPE code for PWR includes three analysis modules, subchannel analysis module, three-dimensional two-phase flow analysis module, and DNB evaluation module. The subchannel module uses drift-flux model to identify the hottest subchannel. The three-dimensional two-phase flow analysis module uses nonhomogeneous and nonequilibrium two fluid model to analyze the detailed three-dimensional two-phase flow behaviors such as void distribution. For DNB heat flux prediction, the DNB evaluation module uses the Weisman model in which is a mechanistic DNB evaluation model. This paper describes the analysis models, analysis techniques and the results of validation by rod bundle test analysis. To date, the average difference between calculated and 11 measured values was -0.6% with a standard deviation of 7.0%. (author)

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Guerre, Catherine; Raquet, Olivier [CEA, DEN/DPC/SCCME/LECA, bat.458, 91191 Gif-sur-Yvette Cedex (France); Duisabeau, Laure [CEA, DEN/DMN/SEMI/LCMI, bat.625, 91191 Gif-sur-Yvette Cedex (France); Turluer, G. [IRSN, DSR/SAMS, BP17, 92262 Fontenay-aux-roses Cedex (France)

    2004-07-01

    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)

  9. 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.

  10. 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)

  11. 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.

  12. Mutated primer binding sites interacting with different tRNAs allow efficient murine leukemia virus replication

    DEFF Research Database (Denmark)

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

    1993-01-01

    Two Akv murine leukemia virus-based retroviral vectors with primer binding sites matching tRNA(Gln-1) and tRNA(Lys-3) were constructed. The transduction efficiency of these mutated vectors was found to be comparable to that of a vector carrying the wild-type primer binding site matching t......RNA(Pro). Polymerase chain reaction amplification and sequence analysis of transduced proviruses confirmed the transfer of vectors with mutated primer binding sites and further showed that tRNA(Gln-2) may act efficiently in conjunction with the tRNA(Gln-1) primer binding site. We conclude that murine leukemia virus...... can replicate by using various tRNA molecules as primers and propose primer binding site-tRNA primer interactions to be of major importance for tRNA primer selection. However, efficient primer selection does not require perfect Watson-Crick base pairing at all 18 positions of the primer binding site....

  13. Two-phase flow measurement in the upper plenum of a PWR during reflood

    International Nuclear Information System (INIS)

    In order to study the two-phase hydrodynamics of a pressurized water reactor during emergency reflood, a group of four instruments has been developed for installation at the core and upper plenum interface. These instruments are a drag body, a cut-out portion of the end-box tie plate; a turbine meter located above one of the tie-plate holes; a differential pressure measurement across the tie plate; and a collapsed liquid level measurement above the tie plate. A single-module air and water apparatus simulates the PWR during reflood. Tests are done with the upper plenum empty, and with it containing simulated control rod guide tubes and baffle plates. The flow regimes studied are flooded and unflooded countercurrent flow, upflow, and combined injection. The upper plenum internals have a significant effect on the behavior of the two-phase flow. An algorithm to determine the individual mass flowrates of gas and liquid is presented. This algorithm depends on the drag body force measurement, the differential pressure measurement, the collapsed liquid level, and the flooding line

  14. 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. ...

  15. 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

  16. 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

  17. 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

  18. Fatigue crack growth threshold of austenitic stainless steels in simulated PWR primary water

    International Nuclear Information System (INIS)

    Many studies have revealed that fatigue crack growth (FCG) rate of austenitic stainless steels is accelerated in light water reactor environment compared to that in air at room temperature. Major driving factors in the acceleration of FCG rate are stress ratio, temperature and stress rise time. Based on this knowledge, FCG curves have been developed considering these factors as parameters. However, there are few data of FCG threshold ΔKth in light water reactor environment. Hence it is necessary to clarify FCG rate under near-threshold condition for more accurate evaluation of fatigue crack growth behavior under cyclic stress with relatively low ΔK. In the present study, therefore, ΔKth was determined for austenitic stainless steels in simulated PWR primary water, and FCG behavior under near-threshold condition was revealed by collecting fatigue crack propagation data. The results are summarized as follows: No propagation of fatigue crack was found in high temperature water, and there was a definite ΔKth. Average ΔKeff,th was 4.3 MPa·m0.5 at 325degC, 3.3 MPa·m0.5 at 100degC, and there was no considerable reduction compared to currently known ΔKeff,th in air. Thus, it was revealed tha ambient conditions had minimal effect, on ΔKeff,th, ΔKth increases with increasing temperature and decreasing frequency. As a result of fracture surface observation, oxide-induced-crack-closure was considered to be a cause of the dependency described above. In addition, it was suggested that changes in material properties also had influence on ΔKth, since ΔKeff,th itself increased at elevated temperature. (author)

  19. Impact forces on a core shroud of an excited PWR fuel assembly

    International Nuclear Information System (INIS)

    Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment

  20. 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

  1. A primer on wavelets and their scientific applications

    CERN Document Server

    Walker, James S

    2008-01-01

    In the first edition of his seminal introduction to wavelets, James S. Walker informed us that the potential applications for wavelets were virtually unlimited. Since that time thousands of published papers have proven him true, while also necessitating the creation of a new edition of his bestselling primer. Updated and fully revised to include the latest developments, this second edition of A Primer on Wavelets and Their Scientific Applications guides readers through the main ideas of wavelet analysis in order to develop a thorough appreciation of wavelet applications. Ingeniously relying o

  2. 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

  3. 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.

  4. Thermal-hydraulic model verification calculation of PWR tests

    International Nuclear Information System (INIS)

    Test PWR 5 determined forces and pressure differences across the reactor pressure vessel and the internals in the test vessel during the first 80 ms by means of the present test data tapes. Furthermore a comparison between the measured data and those determined with the aid of the LECK program system is carried out. The following results were obtained in this connection: The qualitative pattern as compared between calculation and measurement shows a good agreement. Higher pressure differences resulted across the components due to the higher pressure gradients in the initial phase of the blowdown verification in the calculations. The best agreement of the pressure gradients was obtained with the verification calculations for a rupture opening time of 6 ms. Since there was no fluid/structural-dynamic coupling it was not possible to simulate the premature pressure reduction within the core barrel. The distribution of the initial temperature in the calculation did not always agree with that during the test. (orig.)

  5. 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

  6. Optimization of chemistry in PWR and VVER nuclear power plants

    International Nuclear Information System (INIS)

    This paper, based on international feedback and studies, proposes potential improvements for PWR and VVER operation: pH optimization in the primary coolant in order to minimize corrosion product transport/deposition and associated radiation exposure, crud induced power shifts (previously called axial offset anomaly), and fuel failure; use of enriched boron acid (enriched with 10B) to easily optimize the above described pH, particularly with the increased use of higher fuel enrichments; zinc addition in the reactor cooling system; establishment of secondary water chemistry specifications which take into consideration the steam generator tubing materials and design to minimize corrosion risk while keeping sufficient plant availability and decreasing environmental impact; amine selection for the secondary system aimed at mitigating steam generator tube fouling, power loss and maintenance costs as well as corrosion risks; overall operating chemistry options designed to minimize environmental impact, such as elimination of condensate polishers and optimum ion exchange resin use. (orig.)

  7. Optimization of chemistry in PWR and VVER Nuclear Power Plants

    International Nuclear Information System (INIS)

    Potential improvements for PWR and VVER operation are proposed: pH optimization in the primary coolant in order to minimize corrosion product transport/deposition and associated dosimetry, Crud Induced Power Shifts (CIPS, previously named AOA), fuel failure; use of EBA (enriched Boron 10) to easily optimize above described pH, particularly with the larger use of higher fuel enrichments; zinc addition in the RCS; secondary water chemistry specifications depending on the steam generator tubing materials and design for minimizing corrosion risk while keeping a sufficient plant availability and decreasing environmental impacts; amine selection for the secondary system for mitigating steam generator tube fouling, power loss and maintenance costs as well as corrosion risks; overall operating chemistry options to minimize environmental impacts, such as elimination of condensate polishers, optimum ion exchange resin use. (N.T.)

  8. 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

  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. 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.

  11. 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)

  12. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    International Nuclear Information System (INIS)

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  13. 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

  14. Ultrasonic Backscattering in Polycrystalline Materials of Pwr Components

    Science.gov (United States)

    Chassignole, B.; Dupond, O.; Fouquet, T.; Rupin, F.

    2011-06-01

    The ultrasonic examination of metallic components of Pressurized Water Reactors (PWR) is an important challenge for the nuclear industry. During the past decades, EDF R&D has undertaken numerous studies in order to improve the NDT process on these applications and to help to their qualification. The present paper deals with the problem of the structural noise which can potentially disturbs the ultrasonic inspection. In particular, this study proposes a modeling approach to simulate the ultrasonic scattering due to coarse grain structures of polycrystalline materials. The methodology is based on the mixing of a grain scale description of the material and a 2D finite element code (ATHENA) developed by EDF to simulate the ultrasonic propagation in isotropic and anisotropic elastic media. The modeling results are compared to experimental acquisitions on mock-ups containing artificial defects.

  15. MAAP/PWR comparisons to other codes and plant data

    International Nuclear Information System (INIS)

    As part of a simulator qualification effort, a cooperating utility assembled an extensive collection of RETRAN and RELAP calculations and plant transient data for a four loop Westinghouse pressurized water reactor (PWR). This information was provided and formed the basis for a relatively complete set of MAAP comparison calculations. The calculations performed include actual loss of offsite power; actual reactor trip; reactor trip calculation with simplified control system operation; failed power operated relief valve; total loss of feedwater with feed and bleed; main steam line break; steam generator tube rupture; steam generator tube rupture with stuck-open steam line PORV; small loss of coolant accident (LOCA); and small local with failure of safety injection systems. The results obtained were fairly good

  16. Review of PWR-related thermal-shock studies

    International Nuclear Information System (INIS)

    Flaw behavior trends associated with pressurized-thermal-shock (PTS) loading of PWR pressure vessels have been under investigation at ORNL for approx.12 years. During that time, eight thermal-shock experiments with thick-walled steel cylinders were conducted as a part of the investigations. These experiments demonstrated, in good agreement with linear elastic fracture mechanics (LEFM), crack initiation and arrest, a series of initiation-arrest events with deep penetration of the wall, long crack jumps without significant dynamic effects at arrest, arrest in a rising K/sub I/ field, extensive surface extension of an initially short and shallow flaw, and warm prestressing with K/sub I/ equal to or less than 0. This information was used in the development of a fracture-mechanics model that is being used extensively in the evaluation of the PTS issue

  17. Shutdown Chemistry Process Development for PWR Primary System

    Energy Technology Data Exchange (ETDEWEB)

    Sung, K.B. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

    1997-12-31

    This study report presents the shutdown chemistry of PWR primary system to reduce and remove the radioactive corrosion products which were deposited on the nuclear fuel rods surface and the outside of core like steam generator channel head, RCS pipings etc. The major research results are the follows ; the deposition radioactive mechanism of corrosion products, the radiochemical composition, the condition of coolant chemistry to promote the dissolution of radioactive cobalt and nickel ferrite, the control method of dissolved hydrogen concentration in the coolant by the mechanical and chemical methods. The another part of study is to investigate the removal characteristics of corrosion product ions and particles by the demineralization system to suggest the method which the system could be operate effectively in shut-down purification period. (author). 19 refs., 25 figs., 48 tabs.

  18. Evaluation of fire probabilistic safety assessment for a PWR plant

    International Nuclear Information System (INIS)

    The internal fire analysis of the level 1 power operation probability safety assessment (PSA) for Maanshan (PWR) Nuclear Power Plant (MNPP) was updated. The fire analysis adopted a scenario-based PSA approach to systematically evaluate fire and smoke hazards and their associated risk impact to MNPP. The result shows that the core damage frequency (CDF) due to fire is about six times lower than the previous one analyzed by the Atomic Energy Council (AEC), Republic of China in 1987. The plant model was modified to reflect the impact of human events and recovery actions during fire. Many tabulated EXCEL spread-sheets were used for evaluation of the fire risk. The fire-induced CDF for MNPP is found to be 2.1 E-6 per year in this study. The relative results of the fire analysis will provide the bases for further risk-informed fire protection evaluation in the near future. (author)

  19. PWR fuel management optimization using continuous particle swarm intelligence

    International Nuclear Information System (INIS)

    The objective of nuclear fuel management is to minimize the cost of electrical energy generation subject to operational and safety constraints. In the present work, a core reload optimization package using continuous version of particle swarm optimization, CRCPSO, which is a combinatorial and discrete one has been developed and mapped on nuclear fuel loading pattern problems. This code is applicable to all types of PWR cores to optimize loading patterns. To evaluate the system, flattening of power inside a WWER-1000 core is considered as an objective function although other variables such as Keff along power peaking factor, burn up and cycle length can be included. Optimization solutions, which improve the safety aspects of a nuclear reactor, may not lead to economical designs. The system performed well in comparison to the developed loading pattern optimizer using Hopfield along SA and GA.

  20. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  1. 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)

  2. Intelligent main control room for advanced PWR plants

    International Nuclear Information System (INIS)

    The design targets of the main control room of nuclear power plants are as follows. (1) To make a good working environment where operators can operate easily. (2) To reduce the work load and operators error. To this end, MHI has been improving main control room design for advanced PWR plants. The new intelligent main control room consists of a soft operation console and a large display panel. According to our evaluation, the work load and human error of the new main control room are reduced by about 35% compared with the latest plants. This new design will be used to plan new plants and will have the additional feature of saving costs by standardizing plant design. (author)

  3. Conceptual design of SPWR, a PWR with enhanced passive safety

    International Nuclear Information System (INIS)

    A conceptual design has been carried out on a new type of integrated pressurized water reactor, SPWR (System-integrated PWR). This reactor installs a poison tank (borated water filled) in the reactor vessel instead of control rod drive system. Three hydraulic pressure valves are installed as the upper interface between the poison tank and primary coolant. A 700MWe power plant with twin 1100MWt SPWRs which are installed in a reactor building has been studied. Design and analysis have been made on the reactor core, reactor (reactor vessel, steam generator, main circulating pump, pressurizer, poison tank and their integration), plant systems (main and sub systems), layout, construction scheme, operation and maintenance, safety related components, reactor dynamics, economics and R and D needs. Passive safety features are also studied. (author)

  4. Residual heat removal in a PWR using a passive system

    International Nuclear Information System (INIS)

    The present work is made in the frame of the studies that are performed at the French Atomic Energy Commission on the innovative safety systems. The system which is discussed here is devoted to the residual heat removal. It can be used for a current french 3 loops PWR in place of the combination auxiliary feedwater system - atmospheric relief valve. A blackout transient, without auxiliary feedwater, is calculated, using the CATHARE code, in order to assess the capabilities of the system. Some complementary scenarios are calculated, assuming the intervention of other systems after a while, for example restart of the primary pumps and manual opening of the atmospheric relief valves. The influence of non condensable gases is also discussed. 7 refs., 17 figs

  5. Fuzzy logic control for PWR load-follow

    International Nuclear Information System (INIS)

    The developments of fuzzy logic control theory promote the application of fuzzy logic controller to load-follow in Pressurized Water Reactors. A control law combined fuzzy logic controller to load-follow in Pressurized Water Reactors. A control law combined fuzzy logic controller with conventional PID controller, using the strategy of output gains varying with nuclear reactor power, was proposed to control load-follow operations in PWR. This method solves the nonlinear time-varying close-loop control problem and overcomes the shortcomings and limitations of the model-based method. The simulation results show the method is of both satisfactory dynamic performance and high steady state precision. This approach will improve the automaticity of load-follow operations

  6. Development of a general nodalization scheme for PWR simulators

    International Nuclear Information System (INIS)

    The paper deals with the development of four nodalizations of PWR simulators for Cathare 2 V1.3E code. The nodalizations have been set up using the same general scheme for the considered facilities (Lobi, Spes, Bethsy, Lstf). The geometrical configuration of the various plants considered has been reproduced representing the different zones with the same elements in the code. Criteria already tested for nodalizations development have been followed to assure the geometrical fidelity to the represented systems and new criteria have been introduced to assure the maximum possible similarity among the nodalizations. This activity will lead to reduce the effect of differences in the nodalization when comparing calculations of similar experiments, in particular counterpart tests performed in differently scaled facilities. The nodalizations that have been set up are suitable for every kind of transient. The four nodalizations have been tested at a steady state level against experimental data derived from the facilities. (author)

  7. Numerical regulation of a test facility of materials for PWR

    International Nuclear Information System (INIS)

    The installation aims at testing materials used in nuclear power plants; tests consists in simulations of a design basis accident (failure of a primary circuit of a PWR type reactor) for a qualification of these materials. A description of the test installation, of the thermodynamic control, and of the control system is presented. The organisation of the software is then given: description of the sequence chaining monitor, operation, list and function of the programs. The analog information processing is also presented (data transmission). A real-time microcomputer and clock are used for this work. The microprocessor is the 6800 of MOTOROLA. The microcomputer used has been built around the MC 6800; its structure is described. The data acquisition include an analog data acquisition system and a numerical data acquisition system. Laboratory and on-site tests are finally presented

  8. 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)

  9. DEPCO-MULTI, Subcooled Decompression in PWR Primary System LOCA

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: DEPCO-MULTI is used to analyze the subcooled decompression in a multiple pipe network such as a PWR primary coolant system after an instantaneous pipe break. 2 - Method of solution: The unsteady one dimensional mass, momentum and energy conservation laws are converted to a set of compatibility conditions and characteristic directions equivalent to them. After the simplification of the compatibility conditions, they are numerically integrated by the explicit formula. 3 - Restrictions on the complexity of the problem: (1) Maximum number of different thermodynamic initial conditions: 5; (2) Maximum output points of: - pressure history: 5; - pressure difference history: 3; - flow velocity history: 1. (3) The assumptions of thermal equilibrium, no buoyancy effect, no heat transfer and one-dimensional analysis are used

  10. 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.

  11. 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

  12. 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)

  13. PWR power plant pump reliability data. Interim report

    International Nuclear Information System (INIS)

    This report represents a portion of the EPRI effort to collect data relevant for use in nuclear power plant risk studies. The work reported here involved collection and analysis of failure and repair data for pumps in PWR plant operating and safety systems. Failures are classified by failure mode, cause, and part. The failure parameters estimated for pumps are the probability of failure on demand and rate of failure per unit of running time. Different approaches to aggregating the data are presented, and failure parameters are estimated under each approach. Common cause failures are identified, and beta factors are calculated for pumps in some systems. Repair man-hours data are used to estimate a repair time distribution, and the unavailability of pumps in different systems due to repair after failure is calculated

  14. Toward an early detection of PWR control rod anomalous dropping

    International Nuclear Information System (INIS)

    Some anomalous PWR control rods dropping occurred in the past. It is assumed to be caused by a geometrical deformation of its guide tube, which might be related with neutron fluence and its sharp changes. Now at days, this problem is an open field of research, oriented to the understanding and prevention of the event. Work here is focused toward early detection. A differential equation modelling control rod free fall movement is found. There result three acceleration terms: gravity; friction with fluid; and friction with its guide tube. From recorded Plant measurements, both friction coefficients are estimated. The one from guide tube experiences a large variation in case of anomalous dropping; so relationship with neutron fluence is proposed for the prevention purpose. (Author)

  15. Influence of spectral history on PWR full core calculation results

    International Nuclear Information System (INIS)

    The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect-a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. A cross section correction method based on Pu-239 concentration was implemented in the reactor dynamic code DYN3D. This paper describes the influence of a cross section correction on full-core calculation results. Steady-state and burnup characteristics of a PWR equilibrium cycle, calculated by DYN3D with and without cross section corrections, are compared. A study has shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. An impact of the correction on transient calculations is studied for a control rod ejection example. (Authors)

  16. Low concentration NP preoxidation condition for PWR decontamination

    International Nuclear Information System (INIS)

    To use preoxidation condition with low concentration NP (nitric acid permanganate) instead of conventional high concentration AP (alkline permanganate ) for PWR oxidation decontamination (POD) was summarized. Experiments including three parts have been performed. The defilming performance and decontamination factor of preoxidation with low concentration NP, which is 100, 10 times lower than that of AP are better than that with high concentration AP. The reason has been studied with the aid of prefilmed specimens of corrosion potential measuring in NP solution and chromium release in NP and AP solutions. The behaviour of alloy 13 prefilmed specimen in NP preoxidation solution is different from 18-8 ss and Incoloy 800. In the low acidity, the corrosion potential moves toward positive direction as the acidity becomes high

  17. Delayed phenomena analysis from French PWR containment instrumentation system

    International Nuclear Information System (INIS)

    The analysis of the large amount of measurements which has been now gathered by EDF on its twenty two PWR 900 MW shows that the behaviour of concrete under creep and shrinkage effects is in good agreement with the values given as correct estimates by french regulations and taken into account for the design of nuclear prestressed structures. None of the containment buildings studied here showed significant differences with the regulations theoretical values and consequently all the measurements remain in the field of the allowable strain variations used for design. On the other hand, if the instant loading elastic modulus is clearly determined for each containment, and its effect on theoretical creep taken into account, it was not possible up till now to extract from measurements some particular effects such as type of concrete and agregates or climatic effects. (orig.)

  18. 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

  19. Analysis of reactivity insertion accidents in PWR reactors

    International Nuclear Information System (INIS)

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author)

  20. 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)

  1. Corrosion Resistance Evaluation of HANA Claddings in Commercial PWR

    International Nuclear Information System (INIS)

    Korea Atomic Energy Research Institute (KAERI) in collaboration with KEPCO Nuclear Fuel (KNF) developed newly-advanced alloy which are named HANA (High-performance Alloy for Nuclear Application) for high burnup PWR nuclear fuel, showed an excellent out-pile corrosion resistance in PWR simulating loop conditions. And in-pile corrosion resistance of HANA claddings, which was examined at the first provisional inspection after -185 FPD of irradiation in the Halden Reactor, and also shown superior to the other references alloy. Also, other researches showed a much better corrosion resistance when compared to the other Zr-based alloy in various corrosion conditions. In this study, the LTA program for newly-developed fuel assembly (HIPER) with the HANA claddings was implemented to justify the performance for 3 cycles of operation schedule in Hanul nuclear power plant. The objective of this study is to compare corrosion properties of reference alloy with HANA claddings loaded in Hanul nuclear power plant.. For the examination procedures, the oxide thickness measurements method and equipment of PSE are described in detail as follow in measurement methods chapter. Finally, based on the above mentioned measurements method, the summarized oxide thickness data obtained from PSE are evaluated for the corrosion resistance in commercial nuclear power plant and some discussion for the corrosion resistance are described. In the past, corrosion resistance of HANA claddings was successfully conducted in test reactor. In this study, the corrosion characteristic of HANA claddings which are applied to HIPER is examined in the commercial nuclear power plant. HANA claddings in the HIPER showed a more improved corrosion resistance than reference alloy claddings and are evaluated well with meeting the oxide thickness criteria

  2. 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

  3. 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

  4. PWR operation and reloading: EDF experience and developments

    International Nuclear Information System (INIS)

    The large experience accumulated by EDF in PWR operation and reloading for about fifteen years required reliable and industrial techniques. Presently, about 54 units of 900 MWe and 1300 MWe PWR's are being operated through various fuel managements (three-batch cycle, four-batch cycle, plutonium recycling). EDF has developed two sets of automatized computational sequences with automatic generation of input data and core calculations for both, the Loading Pattern (LP) optimization and initialization of input data (fuel reshuffling), and for reload related calculations (safety evaluation, start-up physics tests prediction, operating data). As far as the LP search is concerned, it consists in a technique of 'trial and error' based upon knowledge and which is under very severe constraints. Then, reload values prediction and core following are performed with codes and calculational methods which have a high level of qualification and calibration over the large experience of in-core measurements. With respect to these different points, continuous efforts are done aimed at improving the overall reloading methods. Developments are being achieved at different levels. Because of load following perturbations, on-line and off-line core power distribution followings are evaluated with fast nodal CAROLINE code. This one is derived from the 3D design COCCINELLE code developed by EDF, and whose main features are 3D core calculations with optimized numerical schemes and fast resolution techniques, fuel thermal and neutronic feed-back effects modelling (pin by pin). As an alternative to LP manual design used currently, EDF has examined two possible approaches: expert system and optimization package. As far as automatic sequences are concerned, a new technique of automatic generation of input files was evaluated but priority has been given to improvements in physics by more 3D extensive calculations with the new COCCINELLE code

  5. 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)

  6. 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)

  7. Integrated chemical effects test program for PWR sump performance assessment

    International Nuclear Information System (INIS)

    Products attributable to chemical interactions between the emergency core cooling system (ECCS) containment spray water and exposed materials (such as metal surfaces, paint chips, and fiberglass insulation debris) could impede the performance of ECCS recirculation following a loss-of-coolant accident (LOCA) at a pressurized-water reactor (PWR). Five tests have been conducted in the ICET (Integrated Chemical Effects Test) test loop in order to simulate the chemical environment present inside a PWR containment water pool following a LOCA. The tests were conducted for 30 days at a constant temperature of 60 Celsius degrees. The materials tested within this environment included representative amounts of submerged and un-submerged aluminum, copper, concrete, zinc, carbon steel, and insulation samples (either 100% fiberglass or a combination of 80% calcium-silicate and 20% fiberglass by volume). Representative amounts of concrete dust and latent debris were also added to the test solution. Water was circulated through the bottom portion of the test tank during the entire test to achieve representative flow rates over the submerged specimens. Overall, the ICET program provided some insights and initial understanding regarding solution chemistry, as well as the types and amounts of chemical reaction products that may form in the ECCS containment sump pool. The observed chemical products may potentially contribute to pressure losses across a debris-laden sump screen, as well as performance degradation of ECCS components downstream of the sump screen. The ICET results indicate that: -1) chemical reaction products with varied quantities, consistencies, attributes, and apparent formation mechanisms were found; -2) containment materials (metallic, non-metallic, and insulation debris), pH, buffering agent, temperature, and time are all important variables that influence chemical product formation; and -3) changes to one important environmental variable (e.g., pH adjusting agent

  8. PWR operation and reloading: EDF experience and developments

    International Nuclear Information System (INIS)

    The large experience accumulated by EDF in PWR operation and reloading for about fifteen years required reliable and industrial techniques. Presently, about 54 units of 900 MWe and 1300 MWe PWR's are being operated through various fuel management (three-batch cycle, four-batch cycle, plutonium recycling). EDF has developed two sets of automatized computational sequences with automatic generation of input data and core calculations for both, the Loading Pattern (LP) optimization and initialization of input data (fuel reshuffling), and for reload related calculations (safety evaluation, start-up physics tests prediction, operating data). As far as the LP search is concerned, it consists in a technique of ''trial and error' based upon knowledge and which is under very severe constraints. Then, reload values prediction and core following are performed with codes and calculational methods which have a high level of qualification and calibration over the large experience of in-core measurements. With respect to these different points, continuous efforts are done aimed at improving the overall reloading methods. Developments are being achieved at different levels. Because of load following perturbations, on-line and off-line core power distribution followings are evaluated with fast nodal CAROLINE code. This one is derived from the 3D design COCCINELLE code developed by EDF, and whose main features are 3D core calculations with optimized numerical schemes and fast resolution techniques, fuel thermal and neutronic feed-back effects modelling (pin by pin). As an alternative to LP manual design used currently, EDF has examined two possible approaches: Expert system and optimization package. As far as automatic sequences are concerned, a new technique of automatic generation of input files was evaluated but priority has been given to improvements in physics by more 3D extensive calculations with the new COCCINELLE code. (author). 4 refs, 3 figs

  9. 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

  10. Presentations on Ageing management of PWR piping systems

    International Nuclear Information System (INIS)

    The structural integrity of the reactor coolant system (RCS) of PWR's is a key safety issue. As high-energy piping and essential piping system for cooling the reactor core, the design basis considered different hypothetical double end guillotine break ruptures, generally in 11 locations [fig. 1]. The consequences of that importance are the design, fabrication rules and the surveillance programs of these lines in operation. The field experience is in accordance with these precautions and limited degradations have been encountered up to now. Different evolution of these initial design bases and different practices of surveillance program are used in a case by case application process in different countries: - leak before break - realistic break opening section and break opening time - risk informed in-service inspection The RCS piping systems of existing PWRs have generally a very high quality standard for design, fabrication and operation rules. The corresponding field experience confirms very limited degradations encountered in these systems. Nevertheless, some degradations appears recently (SCC of DMW in VC SUMMER and RINGHALS), some have no direct consequences encountered (as loss of toughness by thermal ageing), some are potential and not encountered for the moment (but no ISI can justify absence of early degradation due to the thickness of these piping systems); some other questions are not completely covered in this presentation, as risk of brittle fracture for some cladded ferritic pipings. The RCS reliability remains very high, but an important effort has to be maintain to understand, case by case, encountered and potential degradation mechanisms is an essential contribution to assure long term high safety level of PWR plants

  11. 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

  12. 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

  13. 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)

  14. Determination of uranium in PWR spent fuels by coulometric titration method

    International Nuclear Information System (INIS)

    Controlled-potential coulometric titration method was applied in 0.5M sulphuric acid medium for the determination of uranium content in samples of PWR spent fuel. In this study, we discussed some experimental conditions related to the determination of uranium in PWR spent fuel samples. Accuracy(recovery of uranium) for the coulometric determination of 1∼7mg uranium standard was 99.96∼100.88%. Precision(relative standard deviation, rsd) for the coulometric determination(n=3) of 3∼4mg uranium in PWR spent fuel samples was 0.07∼0.68%. Relative error for the results of the potentiometric and coulometric determination of uranium PWR spent fuel samples was +0.65∼-2.76%

  15. Recovery and separation for the trace amounts of iodide in PWR spent fuel

    International Nuclear Information System (INIS)

    An separation and recovery technique for iodide in spent pressurized water reactor (PWR) fuels has been established using a SIMFUEL simulated for spent PWR fuel. The spent PWR fuels were dissolved with mixture of nitric and hydrochloric acids(80; 20 mol%) which can oxidize iodide to iodate through dissolution process. Iodide in uranium matrix and co-exist fission products was separated and recovered by organic extraction of iodine with carbon tetrachloride and by back extraction of iodide with 0.1 M NaHSO3. Recovered iodide was measured using an ion chromatograph/shielding system available for analysis of radioactive materials. In practice, a spent PWR fuel whose burnup rate was 42,261 MWd/MtU was analyzed and then the relation between the burnup and the quantity of the fission products was compared to the calculated by burnup code, Origen 2

  16. Contribution to a model for stress corrosion cracking of Alloy 600 in PWR primary water

    International Nuclear Information System (INIS)

    Nickel base alloys such as Alloy 600 are widely used for Pressurized Water Reactors (PWR) components. One of the main drawbacks of Alloy 600 is its susceptibility to intergranular stress corrosion cracking (IGSCC) in PWR primary water. This phenomenon has been extensively studied since more than 30 years and a lot of data are now available in the literature. However, the models proposed are still under debate as the mechanisms of cracking are still not well-known. The aim of this study is to improve our knowledge of SCC mechanisms of Alloy 600 in PWR primary water. The influence of intergranular carbides precipitation and cold-working on intergranular oxide penetrations after exposure in simulated PWR primary environment was more specifically studied. The morphology, the chemical nature and crystalline structure of the oxide formed at the surface of the samples and inside the grain boundaries were characterized using analytical Transmission Electron Microscopy (TEM). (authors)

  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. 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

    Reverse transcription of retroviral genomes is primed by a tRNA annealed to an 18-nucleotide primer binding site. Here, we present a primer complementation system to study molecular interaction of the replication machinery with the primer and primer binding site in vivo. Introduction of eight base...... substitutions into the primer binding site of a murine leukemia virus-based vector allowed efficient RNA encapsidation but resulted in severely reduced vector replication capacity. Replication was restored upon complementation with a synthetic gene designed to encode a complementary tRNA-like primer, 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...

  19. Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Kao, Lain-Su, E-mail: lskao@iner.gov.tw

    2013-10-15

    Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since T{sub B} is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to

  20. Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

    International Nuclear Information System (INIS)

    Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since TB is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to evaluate the

  1. 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...

  2. PWR plant operator training used full scope simulator incorporated MAAP model

    International Nuclear Information System (INIS)

    NTC makes an effort with the understanding of plant behavior of core damage accident as part of our advanced training. For the Fukushima Daiichi Nuclear Power Station accident, we introduced the MAAP model into PWR operator training full scope simulator and also made the Severe Accident Visual Display unit. From 2014, we will introduce new training program for a core damage accident with PWR operator training full scope simulator incorporated the MAAP model and the Severe Accident Visual Display unit. (author)

  3. 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

  4. 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

  5. Application of PWR LOCA margin with the revised appendix K rule

    International Nuclear Information System (INIS)

    Today's focus for nuclear power plant utility owners is to improve plant performances such that the cost per kilowatthour is minimized with enhanced safety. This paper will discuss the impact of design and licensed margin on PWR plant performance, how these margins can be used to improve PWR performance, and how Westinghouse is addressing the regulatory design limits for large break and small break LOCA which impact core thermal design margin. (orig./GL)

  6. ORIGEN-2 libraries based on JENDL-3.2 for PWR-MOX fuel

    International Nuclear Information System (INIS)

    A set of ORIGEN-2 libraries for PWR MOX fuel was developed based on JENDL-3.2 in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee. The calculational model generating ORIGEN-2 libraries of PWR MOX is explained here in detail. The ORIGEN-2 calculation with the new ORIGEN-2 MOX library can predict the nuclides contents within 10% for U and Pu isotopes and 20% for both minor actinides and main FPs. (author)

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Blood Donation and Transfusion: A Primer for Health Educators.

    Science.gov (United States)

    Felts, W. Michael; Glascoff, Mary A.

    1991-01-01

    Presents a primer for health educators about blood donation and transfusion, examining the nature of human blood, the background of blood transfusion, blood donation criteria, risks related to homologous blood transfusion, directed blood donation, potential alternatives to homologous transfusion, and resources for education on the subject. (SM)

  13. Analysis of cobalt source into the primary coolant system of a PWR

    International Nuclear Information System (INIS)

    The understanding of deposition mechanism of 60Co in primary coolant system is very important to find ways to reduce the radiation worker's exposure. In order to develop a deposition model of 60Co in the primary coolant system, the release rate of cobalt source into the coolant in a PWR should be evaluated. By reviewing previous work regarding 60Co buildup in the primary coolant system of PWR, ionic dissolution of corrosion products from oxide films into the coolant was identified as a governing process of release mechanism of cobalt source into the coolant in PWR condition. Release rate constants, 4.16 x 10-7s-1 for stainless steel and 5.56 x 10-8 s-1 for Inconel-600, were obtained by assuming that the dissolution rate is proportional to the thickness of oxide films in a thin oxide film. Total input of cobalt from structural materials in a typical PWR evaluated from the release rate constants reveals that the main source of cobalt input into the coolant system in the PWR is the corrosion of Inconel-600, which covers more than 65% of total cobalt input the PWR primary coolant system

  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. 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, ... a gas, is a major part of air pollution in cities. When ozone forms air pollution, it's ...

  16. 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, ...

  17. 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)

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

    International Nuclear Information System (INIS)

    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

  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. PWR control rod ejection analysis with the numerical nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Kochunas, B.; Downar, T. J. [Univ. of California at Berkeley, Berkeley (Canada)

    2008-10-15

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of

  1. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then

  2. 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

  3. 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

  4. 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.

  5. 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...

  6. Comparison of gamma dose rate calculations for PWR spent fuel assemblies

    International Nuclear Information System (INIS)

    Gamma dose rate calculations from the U.S. Department of Energy and the French Commissariat a l'Energie Atomique et aux Energies Alternatives were compared for PWR spent fuel assembly models with UO2 fuel at 33 MWd/kg burnup and MOX fuel at 60 MWd/kg. Both sides used their reference physics codes in a three-step calculation procedure involving depleting fresh assemblies to obtain the discharge compositions, obtaining the isotopic gamma source after 30 years of simulated radioactive decay, and applying the source to heterogeneous 3-D transport models to tally the flux at one meter away from the axial midpoint through air. The fluxes were ultimately converted to dose rates with different energy-dependent conversion factors. This study was able to pinpoint the intermediate calculations that exhibited the largest sources of discrepancy between the two approaches. Most notably, the U.S. gamma source calculation accounts for Bremsstrahlung and adjusts its multi-group photon release rates to conserve energy. Despite these differences, very similar dose rates were calculated by both approaches. For the UO2 case, which was intended to benchmark a frequently-cited reference study, both the DOE and CEA calculated 30-year dose rates between 4.6 and 5.8 Sv/h with various conversion factors, which are roughly three times lower than the reference study's results. For the MOX case, the calculated dose rates ranged from 8.5 to 11 Sv/h. Given the same conversion factor, the largest difference between DOE and CEA values was about 1.5 Sv/h. (author)

  7. Aerosols behavior inside a PWR during an accident

    International Nuclear Information System (INIS)

    During very hypothetical accidents occurring in a pressurized water ractor, radioactive aerosols can be released, during core-melt, inside the reactor containment building. A good knowledge of their behavior in the humid containment atmosphere (mass concentration and size distribution) is essential in order to evaluate their harmfulness in case of environment contamination and to design possible filtration devices. Accordingly the Safety Analysis Department of the Atomic Energy Commission uses several computer models, describing the particle formation (BOIL/MARCH), then behavior in the primary circuits (TRAP-MELT), and in the reactor containment building (AEROSOLS-PARFDISEKO-III B). On the one hand, these models have been improved, in particular the one related to the aerosol formation (nature and mass of released particles) using recent experimental results. On the other hand, sensitivity analyses have been performed with the AEROSOLS code which emphasize the particle coagulation parameters: agglomerate shape factors and collision efficiency. Finally, the different computer models have been applied to the study of aerosol behavior during a 900 MWe PWR accident: loss-of-coolant-accident (small break with failure of all safety systems)

  8. 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

  9. 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)

  10. A Feasibility Study of an Integral PWR for Space Applications

    International Nuclear Information System (INIS)

    Fission space power systems are well suited to provide safe, reliable, economic and robust energy sources, in the order of 100 KWe. A preliminary feasibility study of a nuclear fission reactor is here presented with the following requirements: i) high reliability, ii) R and D program of moderate cost, iii) to be deployed within a reasonable period of time (e.g. 2015), iv) to be operated and controlled for a long time (10 years) without human intervention, v) possibly to be also used as a byproduct for some particular terrestrial application (or at least to share common technologies), vi) to start with stationary application. The driving idea is to extend as much as possible the PWR technology, by recurring to an integral type reactor. Two options are evaluated for the electricity production: a Rankine steam cycle and a Rankine organic fluid cycle. The neutronics calculation is based on WIMS code benchmarked with MCNP code. The reactivity control is envisaged by changing the core geometry. The resulting system appears viable and of reasonable size, well fit to the present space vector capabilities. Finally, a set of R and D needs has been identified: cold well, small steam turbines, fluid leakage control, pumps, shielding, steam generator in low-gravity conditions, self pressurizer, control system. A R and D program of reasonable extent may yield the needed answers, and some demanding researches are of interest for the new generation Light Water Reactors. (authors)

  11. 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)

  12. 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

  13. 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.

  14. 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

  15. Assessment of spectral history influence on PWR and WWER core

    International Nuclear Information System (INIS)

    The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect - a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. Neglecting this effect leads to an additional component of error in neutron-physical characteristics. Two solution approaches to this problem implemented in the reactor dynamic code DYN3D are described and compared in this paper: a cross section correction method based on 239Pu concentration and separate cross sections treatment for each axial layer of reactor core. Steady-state and burnup characteristics of a PWR and a WWER-1000 cores, calculated by DYN3D with and without cross section corrections, are compared. An impact of the correction on transient calculations is studied for a control rod ejection example. Studies have shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. Two different correction methods have shown similar major effects. (orig.)

  16. Experiments on the load following behaviour of PWR fuel rods

    International Nuclear Information System (INIS)

    KWU had studied the effects of load following operation on fuel performance from the beginning of commercial operation of nuclear power plants: The first power cycling experiments were started in 1970 in the nuclear power plant Obrigheim (KWO) and in the High Flux Reactor (HFR) Petten. These power cycling tests performed at various power levels and burnups of up to 25 GWd/t(U) showed that the fuel rod cycling performance compares well with the performance of fuel rods operated under essentially constant load at comparable power levels. Two additional cycling tests as described in this paper were performed on the HFR Petten with preirradiated PWR fuel rods having burnups of up to 40 GWd/t(U). These experiments comprised up to 60 cycles between 250/360 W/cm and 215/320 W/cm with 10% power overshoot (400, 370 W/cm) after each cycle. Also, these experiments ended up with sound fuel rods. Moreover, detailed investigations before and between power cycles and after experiment termination showed clearly that the fuel performance corresponds to a single ramp to peak power and that the cycling effects are indeed very small. This confirmed earlier findings that due to crack reversal in the UO2 the cyclic dimensional changes mainly occur in the UO2 itself. Altogether the experiments show that power cycling does not lead to fuel rod failures, which is also confirmed by successful load follow operation in commercial power plants. (orig.)

  17. On thermodynamic advantages of hybrid PWR-desalination plants

    Energy Technology Data Exchange (ETDEWEB)

    Ansari, K.; Sayyaadi, H.; Amidpour, M.; Saffari, A.; Sabzaligoll, T. [Univ. of Technology, Tehran (Iran, Islamic Republic of). Dept. of Mechanical Engineering

    2008-07-01

    Nuclear desalination processes are used to produce power and potable water, and are regarded as more thermodynamically efficient and economically feasible than single purpose nuclear generators and water production plants. This study discussed a 1000 MW PWR nuclear power plant combined with a MED-TVC desalination unit with a capacity of 25,000 m{sup 3} per day. The dual purpose plant consisted of 3 interconnected systems, notably (1) a nuclear power plant with a conversion cycle for steam power generation and a turbo generator connection, (2) a coupling system, and (3) a thermal seawater desalination plant. Exergetic simulations were conducted to obtained energy and exergy flows for the hybrid plant. Exergetic efficiency, exergy destructions, and exergy losses were obtained for the proposed plant. Results of the analysis indicated that major exergy destructions occurred within the nuclear reactor. Turbines and steam generators were other sources of exergy destruction. It was concluded that the desalination unit was only responsible for 1.1 per cent of the total exergy destruction of the hybrid plant. 16 refs., 6 tabs., 5 figs.

  18. Localization and manufacturing technology of materials for PWR plants

    International Nuclear Information System (INIS)

    The primary coolant system of PWR type reactor consists of reactor vessel, steam generator, pressurizer, primary coolant piping, and primary coolant pump. In the case of forged metal all of required materials used in above mentioned system are being produced in Korea Heavy Industries and Construction Co.. Small quantities of raw materials of austenite series stainless rolling mill products, primary tubes, and heat transfer pipings of steam generators were imported and manufactured domestically. But the primary coolant pumps are directly imported. Structural materials being installed inside reactor vessel and control element drive mechanisms are being designed and manufactured by Korea Heavy Industries and Construction Co. from Yonggwang 5 reactor. Indigenous production of rotor, bucket, nozzle, casing, bolt, and valve was totally accomplished. The motives of indigenous production of products were the results of continued investment on research activities and equipment and instruments. The materials used in primary coolant system were produced and manufactured only by the companies which holds a stringent quality control programs and has ASME MO Stamp. But the required quantity of materials from those companies is so small that most of them are imported. Other than that almost all of materials for nuclear power reactor are domestically produced. (Hong, J. S.)

  19. 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)

  20. Nature and behaviour of particulates in PWR coolants

    International Nuclear Information System (INIS)

    Corrosion product species transported by PWR coolants are present in both soluble and insoluble form. Whereas many comparative studies of corrosion products and their activated species refer to the total concentration carried by the coolant, few specifically address the nature and behavior of the insoluble component. The information summarised here is from five Belgian PWRs where continuous-flow capillary samplers were installed as a modification to installed coolant sampling facilities. A series of sampling campaigns were undertaken covering all phases of reactor operations and transient conditions. Particulate populations can vary widely from reactor to reactor and also under steady operational conditions in the same system. Some variation of particle size is apparent with reactor age. During commissioning nickel-rich particles were dominant whereas iron was the major constituent during power operation. It was found that filterable material above 0.1 μm in size could account for between 50 and 90% of total coolant borne activity due to cobalt-58 and 60. Other characteristics of coolant particulates are considered. (author)