999 resultados para Pressurized water reactors.


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"June 1959."

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"Including errata of March 7, 1960 and Supplement No. 1, Revised March 11, 1960."

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A total of 193 annotated references to unclassified reports on the design, development and construction of the Shippingport Pressurized Water Reactor is presented. Author, subject, and report number indexes are included.

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From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.

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"June 1959."

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Reproduced by NTIS.

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Prepared under Contract AT(04-3)-165 for the U.S. Atomic Energy Commission, San Francisco Operations Office.

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Prepared under Contract AT(04-3)-165 for the U.S. Atomic Energy Commission, San Francisco Operations Office.

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"April 1981" (v. 1); "June 1981" (v. 2)

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The investigation of insulation debris generation, transport, and sedimentation becomes more important with regard to reactor safety research for pressurized water reactors and boiling water reactors when considering the long-term behavior of emergency core coolant systems during all types of loss-of-coolant accidents (LOCAs). The insulation debris released near the break during a LOCA incident consists of a mixture of disparate particle populations that varies with size, shape, consistency, and other properties. Some fractions of the released insulation debris can be transported into the reactor sump, where it may perturb/impinge on the emergency core cooling systems. Open questions of generic interest are, for example, the particle load on strainers and corresponding pressure drop, the sedimentation of the insulation debris in a water pool, and its possible resuspension and transport in the sump water flow. A joint research project on such questions is being performed in cooperation with the University of Applied Sciences Zittau/Görlitz. The project deals with the experimental investigation and the development of computational fluid dynamics (CFD) models for the description of particle transport phenomena in coolant flow. While the experiments are performed at the University of Applied Sciences Zittau/Görlitz, the theoretical work is concentrated at Forschungszentrum Dresden-Rossendorf. In the current paper the basic concepts for CFD modeling are described and feasibility studies including the conceptual design of the experiments are presented.

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Diplomityössä selvitettiin Fortum Power and Heat Oy:n Loviisan VVER-440 painevesireaktorilaitosten termisen tehon laskentaan liittyviä epävarmuuksia. Laitoksen turvallisuusteknisissä käyttöehdoissa (TTKE) määrätään reaktorin suurimmaksi sallituksi lämpötehoksi 1500 MW. Tähän perustuen haluttiin selvittää nykyiseen RT1 laskentaan liittyvät epävarmuudet tarkastamalla nykyinen laskenta ja siinä käytetyt termohydrauliset laskentasovitteet. Työn alussa selostetaan lyhyesti Loviisan voimalaitoksen toimintaperiaate, jonka jälkeen esitellään laskentaan osallistuvat prosessimittaukset ja niihin liittyvät epävarmuustekijät. Mittauksille määritettiin epävarmuudet käyttäen hyödyksi komponenttivalmistajien tietoja sekä laitoksen kalibrointitodistuksia ja näiden lisäksi laskettiin standardin mukainen virhe virtauslaipoille. Edellä mainittujen virheiden perusteella voitiin laskea tehon epävarmuudet yksittäiselle höyrystimelle, josta edelleen varianssien summamenetelmällä saatiin reaktorin termiselle teholle 0,78 %:n epävarmuus 95 % luottamustasolla. Laskettua tehon epävarmuutta verrattiin Monte Carlo -menetelmällä suoritettuun tarkistuslaskentaan, jolla termisen tehon epävarmuudeksi saatiin 0,53 %, luottamustason ollessa 95 %. Työssä tarkasteltiin keskiarvotuksen vaikutusta mittausdataan. Näissä tarkasteluissa havaittiin pinnansäädöstä aiheutuva reaktoritehon huojunta, joka oli työn merkittävin havainto.