Particular intervention plan of The Bugey Nuclear Power Plant

International Nuclear Information System (INIS)

2014-01-01

The Particular intervention plan (PPI in French) is an emergency plan which foresees the measures and means to be implemented to address the potential risks of the presence and operation of a nuclear facility. This plan is implemented and developed by the Prefect in case of nuclear accident (or incident leading to a potential accident), the impact of which extending beyond the facility perimeter. It represents a special section of the organisation plan for civil protection response (ORSEC plan). The PPI foresees the necessary measures and means for crisis management during the first hours following the accident and is triggered by the Department Prefect according to the information provided by the facility operator. Its aim is to protect the populations leaving within 10 km of the facility against a potential radiological hazard. The PPI describes: the facility, the intervention area, the protection measures for the population, the conditions of emergency plan triggering, the crisis organisation, the action forms of the different services, and the post-accident stage. This document is the public version of the Particular intervention plan of the Bugey NPP (Ain, France)

Integrated thermal analysis of top-shield and reactor vault of Indian FBR-600

International Nuclear Information System (INIS)

Rajendrakumar, M.; Velusamy, K.; Selvaraj, P.

2015-01-01

The design for next generation fast breeder reactors (FBR-600) has been commenced with enhanced safety and improved economy as the main targets. The Top Shield (TS) of Prototype Fast Breeder Reactor (PFBR) is a box type structure consisting of Roof Slab (RS), Small Rotatable Plug (SRP), and Large Rotatable Plug (LRP). The large box type structure with many penetrations posed difficulties during manufacturing. Because of the required high load carrying capabilities, a dome shaped thick plate roof slab is conceived for FBR-600. Main Vessel (MV) which holds the primary sodium and associated components is welded to the RS through a triple joint. Reactor vault (RV) is a thick concrete structure which supports MV and Safety Vessel (SV). The temperature of RV concrete has to be less than 338 K (65°C) under normal operating heat loads (full and part load conditions) and less than 363 K (90°C) under Safety Grade Decay Heat Removal (SGDHR) conditions with one cooling loop in service. The temperature in the component penetrations of the RS should be greater than 120°C to avoid sodium aerosol deposition. Similarly, the temperature of the LRP and SRP has to be ∼120°C to protect the elastomeric seals provided to these structures. Further, the heat load to RV transferred by direct conduction by roof slab support has to be minimum. To meet these conflicting thermal requirements, detailed multi-physics CFD calculations have been performed to finalize, (i) the insulation requirements on the top of roof slab, (ii) number and position of reflective insulation plates below the bottom plate of roof slab/rotating plugs, (iii) air flow rate for various zones of the top shield and (iv) water flow rate and pitch of water cooling pipes for the reactor vault. (author)

3SE: Expert system for the survey of electric sources - CPN of Bugey

International Nuclear Information System (INIS)

Ancelin, J.; Cheriaux, F.; Drelon, R.; Gaussot, J.P.; Marion, B.; Maurin, S.; Pichot, D.; Sancerni, G.; Voisin, C.; Legaud, P.

1990-01-01

The 3SE is an expert system for surveying the electric sources of a 900 MW PWR nuclear power plant. The main objectives of the expert system are: to provide a continuous and a real time support for electric faults data processing; to provide assistance in the electric equipment maintenance; to contribute to the instruction of operators as well as to the data base management of the electric system. Data bases and artificial intelligence techniques are applied. The system's application is based on the accurate knowledge of the nuclear power plant operation and topology, as well as on a model approach. The expert system is applied in the section 2 of the Bugey nuclear power plant. The system which required the effort of 20 engineers x years, is an example of the progress performed in the artificial intelligence field [fr

Activation calculations for dismantling - The feedback of a 7 years experience in activation calculations for graphite gas cooled reactors in France

International Nuclear Information System (INIS)

Eid, M.; Nimal, J.C.; Gerat, L.M.

1994-01-01

This is a revision of the past seven years experience in activation calculations for dismantling. It aims at evaluating the experience and at making better understanding to help in decision making during the following phases. Five gas cooled reactors are shutdown and are waiting for the EDF (Electricite De France) dismantling decision. The sixth (BUGEY1) will be shutdown by 1994 and will be waiting a dismantling decision as well. (authors). 3 figs., 3 tabs

Experiences concerning reactor pressure vessel head penetration inspections; Erfahrungen mit Pruefungen von Reaktordruckbehaelter-Deckeldurchfuehrungen

Energy Technology Data Exchange (ETDEWEB)

Debnar, Angelika [Westinghouse Electric Germany GmbH, Mannheim (Germany)

2009-07-01

Globally observed damage at the control rod drive mechanism nozzles in PWR-type reactors (Bugey-3, Oconee 1,2,3 and ANO-1, David Besse) have triggered enhanced inspection of reactor pressure vessel (RPV) head penetrations. In Germany the regulations require a periodic inspection especially of dissimilar welds. Westinghouse has developed an automated measuring system for RPV heads aimed to inspect welded joints at open nozzles of nozzles with thermosleeves. The testing technology with remote controlled robotics is supposed to perform a weld inspection as complete as possible, restraints result from constructive difficulties for the accessibility. The new gap-scanner DE2008 was qualified at the mock-up and was implemented into the periodic in-service inspection of Neckarwestheim-1.

Investigation of reactivity variations of the Isfahan MNSR reactor due to variations in the thickness of the core top beryllium layer using WIMSD and MCNP codes

Directory of Open Access Journals (Sweden)

A Shirani

2010-12-01

Full Text Available In this work, the Isfahan Miniature Neutron Source Reactor (MNSR is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core and also after 14 years of operation (total operation time of the reactor is calculated. The reactor is then simulated using the MCNP code, and its reactivity variation due to adding a 1.5 mm thick beryllium shim plate to the top of the reactor core, after 7 years of operation, is calculated. The results show good agreement with the available data collected at the revival time. Exess reactivity of the reactor at present time (after 14 years of operation and after 7 years of the the reactor revival time is also determined both experimentally and by calculation, which show good agreement, and indicate that at the present time there is no need to add any further beryllium shim plate to the top of the reactor core. Furthermore, by adding more beryllium layers with various thicknesses to the top of the reactor core, in the input program of the MCNP program, reactivity value of these layers is calculated. From these results, one can predict the necessary beryllium thickness needed to reach a desired reactivity in the MNSR reactor.

Principles of design and construction for the top caps of prestressed concrete reactor pressure vessels

International Nuclear Information System (INIS)

Hughes, A.N.; Bellwood, G.N.; Paton, A.A.

1976-01-01

The building of the top cap poses problems because of the number of penetrations to be cast therein. The fuel and control system routes need to be tightly specified and controlled so that during station life misalignments do not occur which interfere with the fuelling and control operations. The paper outlines the route requirements and illustrates how these affect the tolerances and movements which can be allowed at various stages of construction. Development work is discussed to show the necessity of resolving the different priorities of design, programme and overall pressure vessel construction requirements, so that the reactor build is not inhibited by the special demands of the top cap, and the integration of the monitoring and survey systems during the top cap build are explained. (author)

Proceedings of the Topical meeting on the reactor fuel performance - TopFuel 2012 Transactions

International Nuclear Information System (INIS)

2012-01-01

TopFuel is an annual topical meeting organised by ENS, the American Nuclear Society and the Atomic Energy Society of Japan. TopFuel's primary objective is to bring together leading specialists in the field from around the world to analyse advances in nuclear fuel management technology and to use the findings of the latest cutting-edge research to help manufacture the high performance nuclear fuels of today and tomorrow. Aim is to discuss the challenges facing the developers and manufacturers of new high-performance nuclear fuels - fuels that will help meet current and future energy demand and reduce man's over dependence upon CO 2 -emitting fossil fuels. The technical scope of Top Fuel 2012 includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. (authors)

Thermal design of top shield for PFBR

International Nuclear Information System (INIS)

Gajapathy, R.; Jalaludeen, S.; Selvaraj, A.; Bhoje, S.B.

1988-01-01

India's Liquid Metal Cooled Fast Breeder Reactor programme started with the construction of loop type 13MW(e) Fast Breeder Test Reactor (FBTR) which attained criticality in October 1985. With the experience of FBTR, the design work on pool type 500 MW(e) Prototype Fast Breeder Reactor (PFBR) which will be a forerunner for future commercial fast breeder reactors, has been started. The Top Shield forms the cover for the main vessel which contains the primary circuit. Argon cover gas separates the Top Shield from the free level of hot sodium pool (803K). The Top Shield which is of box type construction consists of control plug, two rotatable plugs and roof slab, assembled together, which provide biological shielding, thermal shielding and leak tight containment at the top of the main vessel. Heat is transferred from the sodium pool to the Top Shield through argon cover gas and through components supported by it and dipped in the sodium pool. The Top Shield should be maintained at the desired operating temperature by incorporating a cooling system inside it. Insulation may be provided below the bottom plate to reduce the heat load to the cooling system, if required. The thermal design of Top Shield consists of estimation of heat transfer to the Top Shield, selection of operating temperature, assessment of insulation requirement, design of cooling system and evaluation of transient temperature changes

Proceedings of the conference on the Safety in Reactor Operations - TopSafe 2012 Transactions

International Nuclear Information System (INIS)

2012-01-01

TopSafe 2012 provides a forum for addressing the current status and future perspectives with regards to safety at nuclear installations worldwide. In view of the on-going discussions and initiatives that have been taken over the last months the European Nuclear Society (ENS) decided organising this edition of this topical conference from 22 to 26 April 2012 in Helsinki, Finland. TopSafe 2012 focus on three main subjects: Safety and related analyses in operating nuclear power plants and other nuclear installations; Safety and Risk Assessment; Trends in nuclear safety for existing and future installations. The conference is directed at a broad range of experts in the area of nuclear safety, including professionals from the different disciplines involved in the safety of nuclear power plants, fuel cycle installations and research reactors. It is aimed at professionals coming from the research organisations, universities, vendors, operators, regulatory bodies as well as policy makers. Top level representatives of the Countries that are constructing new nuclear power plants are invited. Regulators of all individual Countries with nuclear programme are expected to contribute the Conference. (authors)

Improvement of top shield analysis technology for CANDU 6 reactor

International Nuclear Information System (INIS)

Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

1996-07-01

As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

Advanced light water reactor utility requirements document: Volume 1--ALWR policy and summary of top-tier requirements

International Nuclear Information System (INIS)

Anon.

1990-01-01

The U.S. utilities are leading an industry wide effort to establish the technical foundation for the design of the Advanced Light Water Reactor (ALWR). This effort, the ALWR Program, is being managed for the U.S. electric utility industry by the Electric Power Research Institute (EPRI) and includes participation and sponsorship of several international utility companies and close cooperation with the U.S. Department of Energy (DOE). The cornerstone of the ALWR Program is a set of utility design requirements which are contained in the ALWR Requirements Document. The purpose of the Requirement Document is to present a clear, complete statement of utility desires for their next generation of nuclear plants. The Requirements Document covers the entire plant up to the grid interface. It therefore is the basis for an integrated plant design, i.e., nuclear steam supply system and balance of plant, and it emphasizes those areas which are most important to the objective of achieving an ALWR which is excellent with respect to safety, performance, constructibility, and economics. The document applies to both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The Requirements Document is organized in three volumes. Volume 1 summarizes AlWR Program policy statements and top-tier requirements. The top-tier design requirements are categorized by major functions, including safety and investment protection, performance, and design process and constructibility. There is also a set of general design requirements, such as simplification and proven technology, which apply broadly to the ALWR design, and a set of economic goals for the ALWR program. The top-tier design requirements are described further in Volume 1 and are formally invoked as requirements in Volumes 2 and 3

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Use of aquatic mosses for monitoring artificial radionuclides downstream of the nuclear power plant of Bugey (River Rhone, France)

International Nuclear Information System (INIS)

Beaugelin-Seiller, K.; Brottet, D.

1994-01-01

The detection of radionuclides in water, downstream of nuclear installations located on river banks, is often very difficult notably because of their low concentrations. Thus the use of biological indicators is an interesting process to detect radioactive contamination of an aquatic ecosystem. From 1986 to 1990, artificial radionuclides were measured in freshwater mosses sampled downstream of the nuclear power station of Bugey. These field data on the whole, have shown a comparatively good qualitative and quantitative relationship between radioactive composition of liquid waste and radionuclides detected in mosses. In other respects, the results showed up a relatively clear hierarchical structure in the affinity of the different radionuclides for the mosses. To specify these relations, mesh bags containing allochtonous mosses were immersed at four stations downstream of the power plant and regularly sampled during a 10-h waste discharge period. (author)

Removable top nozzle and tool for a nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Wilson, J.F.; Cerni, S.; Gjertsen, R.K.

1986-01-01

A fuel assembly is described for a nuclear reactor including a bottom nozzle, at least one longitudinally extending control rod guide thimble having an upper end and a lower end being attached to the bottom nozzle and projecting upwardly therefrom, transverse grids axially spaced along-the thimble for supporting an array of upstanding fuel rods, and a top nozzle subassembly removable mounted on the upper end of the guide thimble for obtaining top access to the fuel rods upon removal thereof. The top nozzle subassembly consists of: (a) a section integrally formed on the upper end of the guide thimble and having external threads thereon; (b) a lower adapter plate having a guide thimble hole for receiving the guide thimble so as to mount the adapter plate on the guide thimble for slidable movement therealong; (c) a retainer mounted on the guide thimble for restably supporting and limiting the downward movement of the adapter plate along the guide thimble; (d) an upper hold-down plate having a guide thimble passageway with an internal ledge for receiving the thimble so as to mount the hold-down plate on the thimble for slidable movement therealong; (e) spring means interposed between the upper hold-down plate and the lower adapter plate for biasing the hold-down plate upwardly when a downward force is applied thereon whereby the downward force is yieldably transmitted to the fuel assembly; and (f) a collar disposed within the passageway and in abutment with the ledge, the collar having an internal threaded section engageable with the externally threaded section to move the hold-down plate down against the spring means and thereby mounting of the subassembly on the guide thimble

Inception of the light water reactor system in France and its fuel cycle

International Nuclear Information System (INIS)

Gaussot, D.; Elkouby, A.; Sornein, J.

1977-01-01

The electro-nuclear equipment program of Electricite de France (E.D.F.) currently is based on the construction or the commitment of a considerable number of units equiped with pressurized water reactors. The program was preceded by the construction jointly by EDF and Belgian electric utilities of nuclear plants at Chooz (SENA) and Tihange (SEMO). The program as such began with the construction of Units 1 and 2 at Fessenheim, (of which information is given on the start-up), then of Bugey 2, 3, 4 and 5. This was followed by the construction of a series of units of 900 MW, and since 1976, of a series of units of 1300 MW. The successful implementation of such a program is based on a number of technical and organizational guidelines, which are described. The main characteristics of the 3-loop 900 MW and 4-loop 1300 MW NSSS and their fuel are considered. Stress is laid in particular on the development of the 900 MW NSSS from Fessenheim to Bugey. These programs call for winning and processing a great deal of nuclear material right through the fuel cycle. Data are given on the quantities involved and on the production potential permitting the fulfillment of the program (nat. U, enriched UF 6 , fuel subassemblies, reprocessing). The requirements of the EDF, (the NSSS supplier and the industries), the contribution made by the French Atomic Energy Commission (CEA), and international cooperation now in progress, are described. Lastly a number of significant actions both under way and scheduled, are discussed [fr

Reactor Core Internals Replacement of Ikata Units 1 and 2

International Nuclear Information System (INIS)

Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

2012-01-01

This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

The effect of dielectric top lids on materials processing in a low frequency inductively coupled plasma (LF-ICP) reactor

International Nuclear Information System (INIS)

Lim, J.W.M.; Chan, C.S.; Xu, L.; Xu, S.

2014-01-01

The advent of the plasma revolution began in the 1970's with the exploitation of plasma sources for anisotropic etching and processing of materials. In recent years, plasma processing has gained popularity, with research institutions adopting projects in the field and industries implementing dry processing in their production lines. The advantages of utilizing plasma sources would be uniform processing over a large exposed surface area, and the reduction of toxic emissions. This leads to reduced costs borne by manufacturers which could be passed down as consumer savings, and a reduction in negative environmental impacts. Yet, one constraint that plagues the industry would be the control of contaminants in a plasma reactor which becomes evident when reactions are conducted in a clean vacuum environment. In this work, amorphous silicon (a-Si) thin films were grown on glass substrates in a low frequency inductively coupled plasma (LF-ICP) reactor with a top lid made of quartz. Even though the chamber was kept at high vacuum (~10 −4 Pa), it was evident through secondary ion mass spectroscopy (SIMS) and Fourier-transform infra-red spectroscopy (FTIR) that oxygen contaminants were present. With the aid of optical emission spectroscopy (OES) the contaminant species were identified. The design of the LF-ICP reactor was then modified to incorporate an Alumina (Al 2 O 3 ) lid. Results indicate that there were reduced amounts of contaminants present in the reactor, and that an added benefit of increased power transfer to the plasma, improving deposition rate of thin films was realized. The results of this study is conclusive in showing that Al 2 O 3 is a good alternative as a top-lid of an LF-ICP reactor, and offers industries a solution in improving quality and rate of growth of thin films. (author)

Thermal design of top shield

International Nuclear Information System (INIS)

Raghupathy, S.; Velusamy, K.; Parthasarathy, U.; Ghosh, D.; Selvaraj, P.; Chellapandi, P.; Chetal, S.C.

2005-01-01

Full text of publication follows: Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, pool type fast reactor. The top shield forms the top cover for the main vessel (MV) and includes roof slab (RS), large rotatable plug (LRP), small rotatable plug (SRP) and control Plug (CP). RS, LRP and SRP are box type structures consisting of top and bottom plates stiffened by radial stiffeners and vertical penetration shells. TS is exposed to argon cover gas provided above sodium pool on the bottom side and reactor containment building air at the top. Heat transfer takes place through the argon cover gas to the bottom plate of TS. Annular gaps are formed between the components supported on TS and the component penetrations through which cellular convection takes place. A single thermal shield provided below TS reduces the heat flux to the bottom plate to 1.15 kW/m 2 . The MV (SS 316 LN) is welded to RS (carbon steel A48 P2) through a dissimilar metal weld. A step in RS and an anti convection barrier (ACB) outside RS are provided to limit the temperature at the MV-RS junction. The MV is surrounded by safety vessel (SV) and reactor vault made of concrete. Thermal insulation is provided outside SV to limit the heat transfer to the reactor vault. The design requirements of TS are to maintain the operating temperature at 383-393 K, limit the temperature difference (ΔT) across the height of TS to 20 / 100 K under normal operation/loss of cooling, provide minimum annular gap size at the component penetrations, provide a nearly linear temperature gradient in the CP portion within the height of TS, maintain the temperature of top plate of CP > 383 K, limit the ΔT across the top plate of CP to 2 K, limit the temperature near the inflatable / backup seal to 393 K, limit the temperature at the MV-RS junction and the heat flux to the reactor vault. The total heat transferred to TS is estimated to be 210 kW. A dedicated closed loop cooling system with a total flow rate of 10

Electron-neutrino scattering: neutrino magnetic moment and solar neutrino spectrum

International Nuclear Information System (INIS)

Avenier, M.; Bagieu, G.; Barnoux, C.; Bon Nguyen, R.; Brissot, R.; Guerre-Chaley, B.; Laborie, J.M.; Koang, D.H.; Lebrun, D.; Stutz, A.; Vignon, B.

1997-01-01

We have built a low background detector based on a gas time projection chamber surrounded by an active anti-Compton shielding. The detector is installed near a nuclear reactor at Bugey for the experimental study of ν e bar, e - scattering. (authors)

Thermal-hydraulic characteristics of pressurized water reactors during commercial operation. Pt. 5

International Nuclear Information System (INIS)

Procaccia, H.; David, J.; Wazzan, A.R.

1984-01-01

Measured downcomer water and metal shell temperatures in the steam generator No. 20 of the PWR Tricastin 1 show that the downcomer flows is of the swirling type, just as found previously in Bugey 4. A comparison of results for Tricastin 1 and Bugey 4 shows that the addition in Tricastin 1 of a flow distribution baffle plate, between the tube sheet and the first cross plate, while reducing the height of the opening between the tube sheet and the shell surrounding the bundle, may have resulted in the observed reduction (by a factor one half) of sludge deposit upon the tube sheet in Tricastin 1, and in fixing, with extended period of operation, the boiling zone in the cold leg (a desired event) and near the tube-free tube lane. (orig.)

An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

NARCIS (Netherlands)

Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

2009-01-01

In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

Zinc injection on the EDF pressurized light water reactors. Current results and operating experience feedback

International Nuclear Information System (INIS)

Piana, Olivier; Duval, Arnaud; Moleiro, Edgar; Benfarah, Moez; Bretelle, Jean-Luc; Chaigne, Guy

2014-01-01

Nowadays, zinc injection, as well as pH management and hydrogen control, is increasingly considered as an essential element of PWR Primary Water Chemistry worldwide. After a first implementation of zinc injection at Bugey 2 since 2004 and Bugey 4 since 2006, EDF decided to extend this practice, which constitutes a modification of primary circuit chemical conditioning, to other units of its fleet. Currently, 15 among the 58 reactors of the French fleet are injecting depleted zinc acetate into the primary coolant water. Three main goals were identified at the beginning of this program. Indeed, the expected benefits of zinc injection were: Reduction of the rate of generalized corrosion and mitigation of stress corrosion cracking initiation on nickel based alloys (Material goal). Curative or preventive reduction of radiation sources to which workers are exposed (Radiation fields' goal). Mitigation of the AOA or CIPS risks by reduction of corrosion products releases and mitigation of crud deposition (Fuel protection goal). To monitor the zinc addition, EDF has defined a complete survey program concerning: chemistry and radiochemistry responses (primary coolant monitoring of corrosion and fission products and calculation of zinc injected, zinc removed and zinc incorporated in RCS surfaces) ; radiation fields (dose rates and deposited activities measurements) ; materials (statistical analysis of SG tube cracks) ; fuel (oxide thickness measurements and visual exams) ; effluents (corrosion products releases and isotopic distribution follow up) ; wastes (radiochemical characterization of filters). This paper will detail the present results of this monitoring program. It appears that the expected benefits of zinc injection have yet to be fully realized; further operating experience will be required in order to fully evaluate its impact. (author)

Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2009-06-15

SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

International Nuclear Information System (INIS)

2009-06-01

SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing facilities. - 3. Advances in Water

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

ANCCLI Scientific Committee - Contribution to the analysis of modification request file under Article 26 of Decree nr 2007-1557 of the 2 November 2007 related to the releases and water takings on the Bugey site, and of projects of decision of the Nuclear Safety Authority. Study performed at the request of the Bugey CLIS

International Nuclear Information System (INIS)

Chambon, Paul; Sene, Monique

2014-01-01

This report discusses and criticizes some modifications planned on the Bugey site as they have been designed by EDF, commented by the ASN, and perceived by the ANCCLI's scientific committee. These modifications concern: the possibility of implementation of a secondary conditioning with high-pH ethanolamine, the implementation of a processing by pH-controlled massive chlorination for condenser cooling circuits, the implementation of an anti-scale processing of these circuits by injection of polyacrylates and point injection of sulfuric acid (issues related to some bacteria are addressed), the revision of limits of samplings of water, thermal releases and releases of radioactive and chemical substances (comments on the health impact of these releases, discussion for the various concerned chemical species). It also addresses the importance of protocols for a monitoring of the environment

Top-tier requirements for KNGR

International Nuclear Information System (INIS)

Sung-Jae, Ch.; Kwangho, L.; Dong Wook, J.

1996-01-01

In 1992, Korea Electric Power Corporation (KEPCO) has launched the next generation reactor project to develop the standard design of an advanced pressurized water reactor by 2000. This advanced reactor aims to have the sufficient capability to be a safe, environmentally sound and economical energy source for 2000's in Korea. In conjunction with the project development, the program phase I is studied and it is in the Korean Next Generation Reactor (KNGR) first phase project that the requirements of this specification called ''Top-tier'' have been established. These functional requirements are of the first importance for the design, construction and operation of a nuclear power plant. These requirements are divided into safety requirements, serious accidents control, design base requirements, definition of the system characteristics, performance, construction feasibility, economical objectives, site parameters and design processes. The ''Top-tier'' requirements are concentrated on the improvement of the safety and reliability. Safety is one of the first priorities. In particular, the requirements for the design of the next reactors generation must include the capacity to control serious accidents because when an accident occurs, the protection degree is crucial. The KNGR requirements include the existing nuclear power plants competitiveness as well as those of the coal thermal plants. Moreover, when safety is reinforced, the economic competitiveness can be assured. At the present time, a subsequent specification for the KNGR considering the bases of the domestic technology and experimenting the running. (O.M.)

Nuclear reactor fuel assembly with a removable top nozzle

International Nuclear Information System (INIS)

Shallenberger, J.M.; Ferlan, S.J.

1986-01-01

This patent describes a fuel assembly having at least one control rod guide thimble and a top nozzle, the top nozzle including a transversely extending adapter plate. An improved attaching structure is described for removably mounting the top nozzle on the guide thimble comprising: (a) means defining an outer socket in the top nozzle, the outer socket defining means including a passageway extending through the adapter plate and having a first mating element defined in the adapter plate within the passageway; (b) means on an upper end of the guide thimble defining an inner socket, the inner socket defining means including an elongated sleeve having an upper end portion. The upper end portion of the sleeve has a second mating element formed thereon and at least one elongated axial slot defined therein for permitting radial movement of the sleeve upper end portion between a compressed releasing position for removing and inserting the inner socket from and into the outer socket and an expanded locking position for locking the inner and outer sockets together

The Bugey nuclear power plant, at the service of a safe, competitive and CO2-free power generation in the heart of the Rhone-Alpes region

International Nuclear Information System (INIS)

2010-01-01

In less than 20 years, Electricite de France (EDF) has built up a competitive park of 58 nuclear power plants, with no equivalent elsewhere, which represents an installed power of 63.1 GW (85% of EDF's power generation). Inside this nuclear park, the national power generation centre of Bugey comprises 4 production units of 900 MW each (3600 MW as a whole). The facility generated 20.87 billion kWh in 2009, i.e. 5% of the French national power generation and 40% of the energy consumed in the Rhone-Alpes region. This brochure presents the life of the power plant under various aspects: power generation, safety priority and culture, maintenance investments, respect of the environment, long-term fuel and wastes management, local economical involvement, transparency and public information, key figures and dates. (J.S.)

Reduction of the pool-top radiation level in HANARO

International Nuclear Information System (INIS)

Lee, Choong-Sung; Park, Sang-Jun; Kim, Heonil; Park, Yong-Chul; Choi, Young-San

1999-01-01

HANARO is an open-tank-in-pool type reactor. Pool water is the only shielding to minimize the pool top radiation level. During the power ascension test of HANARO, the measured pool top radiation level was higher than the design value because some of the activation products in the coolant reached the pool surface. In order to suppress this rising coolant, the hot water layer system (HWL) was designed and installed to maintain l.2 meter-deep hot water layer whose temperature is 5degC higher than that of the underneath pool surface. After the installation of the HWL system, however, the radiation level of the pool-top did not satisfy the design value. The operation modes of the hot water layer system and the other systems in the reactor pool, which had an effect on the formation of the hot water layer, were changed to reduce pool-top radiation level. After the above efforts, the temperature and the radioactivity distribution in the pool was measured to confirm whether this system blocked the rising coolant. The radiation level at the pool-top was significantly reduced below one tenth of that before installing the HWL and satisfied the design value. It was also confirmed by calculation that this hot water layer system would significantly reduce the release of fission gases to the reactor hall and the environment during the hypothetical accident as well. (author)

Nuclear reactor fuel assembly with a removably top nozzle

International Nuclear Information System (INIS)

Shallenberger, J.M.; Ferlan, S.J.

1985-01-01

The invention relates to a nuclear fuel assembly having an improved attaching structure for removably mounting the top nozzle of the fuel assembly on the upper end of a control-rod guide thimble. The attaching structure comprises an outer socket defined in a portion of the top nozzle, an inner socket extending from the upper end of the guide thimble and removably received in the outer socket for interlocking engagement therewith, and an elongate locking member adapted to be inserted into the inner socket to maintain said interlocking engagement. Removal of the locking member from the inner socket enables the latter to be withdrawn from the outer socket, thereby enabling the top nozzle to be removed from the guide thimble

Prestressed reactor vessel for nuclear power plants

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1982-01-01

With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

High temperature gas cooled nuclear reactor

International Nuclear Information System (INIS)

Hosegood, S.B.; Lockett, G.E.

1975-01-01

For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

Nuclear reactor cavity streaming shield

International Nuclear Information System (INIS)

Klotz, R.J.; Stephen, D.W.

1978-01-01

The upper portion of a nuclear reactor vessel supported in a concrete reactor cavity has a structure mounted below the top of the vessel between the outer vessel wall and the reactor cavity wall which contains hydrogenous material which will attenuate radiation streaming upward between vessel and the reactor cavity wall while preventing pressure buildup during a loss of coolant accident

Technology selection for offshore underwater small modular reactors

International Nuclear Information System (INIS)

Shivan, Koroush; Ballinger, Ronald; Buongiorno, Jacopo; Forsberg, Charles; Kazimi, Mujid; Todreas, Neil

2016-01-01

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 time frame. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO 2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options

Technology selection for offshore underwater small modular reactors

Energy Technology Data Exchange (ETDEWEB)

Shivan, Koroush; Ballinger, Ronald; Buongiorno, Jacopo; Forsberg, Charles; Kazimi, Mujid; Todreas, Neil [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States)

2016-12-15

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 time frame. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO{sub 2} cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

Technology Selection for Offshore Underwater Small Modular Reactors

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-12-01

Full Text Available This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1 a lead–bismuth fast reactor based on the Russian SVBR-100; (2 a novel organic cooled reactor; (3 an innovative superheated water reactor; (4 a boiling water reactor based on Toshiba's LSBWR; and (5 an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80% with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

Atomic pile Directorate, Department of Metallurgy, Departments of Technology, Department of Fuel Elements and Structures, Division of Study of Fuel Elements - Semi annual report on the 1968-10-1

International Nuclear Information System (INIS)

Arnaud, M.; Tortel, J.; Viallet, H.; Marinot, R.; Rulleau, A.; Lestiboudois, G.; Rousseau, G.; Faussat, A.; Ollier, H.; Truffert, J.; Ferrier, C.; Courcon, P.; Rendu, M.; Dieumegard, M.; Bret, A.

1968-01-01

This document gathers a set of reports of studies performed on nuclear fuel elements. The addressed topics are: creep behaviour of UMo and UMoAl tubes and pellets under the action of an external pressure (creep strength of tubes under external pressure, creep strength of pellets under external pressure, uncertainties on irradiation parameters in Pegase), problems related to centring devices (measurements and tests), irradiations of ring elements in power reactors, uranium/sheath metallurgical relationship for Bugey and influence of irradiation (cartridge behaviour in Pegase, long duration irradiation in power reactors, extrapolation in Bugey of results obtained in G2), theoretical study of kinetic oxidation phenomena in metal fuels, tests of leaking cartridges in EdF2, evolution of pressure in EL4 type irradiated fuel rods with ZrCu liners with respect to the conductivity integral, a focus on irradiations of Z0 type fuel elements in Pegase, cluster safety tests with uranium carbide in pile and out of pile, a review of studies performed on fuel elements with blowhole, and application of neutrography to fuel elements

Top closure for control rod drive for nuclear reactor

International Nuclear Information System (INIS)

Raas, J.H.; Schwartz, J.I.

1978-01-01

A removable top closure and venting assembly for the tubular housing of a control rod drive includes a mounting ring threadably inserted in the upper end of the housing, a fluid-sealing closure member beneath the mounting ring and which is mounted in and coupled to the mounting ring by means of a ball and socket joint, a gas vent defined by interconnecting passages extending through the closure and through the ball and socket joint, and a vent valve accessible from the top of the closure assembly. 3 claims, 2 figures

Bugey plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

International Nuclear Information System (INIS)

2011-01-01

This CSA (Complementary Safety Assessment) analyses the robustness of the Bugey plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

Qinshan CANDU project open top construction method

International Nuclear Information System (INIS)

Petrunik, K.J.; Wittann, K.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

2003-01-01

The significant schedule reductions achieved on the Qinshan CANDU Project were due in large part to the incorporation of advanced construction technologies in project design and delivery. For the Qinshan Project, a number of key advantages were realized through the use of the 'Open Top' construction method. This paper discusses the Qinshan Phase III CANDU Project Open Top implementation method. The Open Top method allowed major equipment to be installed simply, via the use of a Very Heavy Lift (VHL) crane and permitted the use of large-scale modularization. The advantages of Open Top construction, such as simplified access, more flexible project scheduling, improved construction safety and quality, and reduced labours are presented. The large-scale modularization of the Reactor Building Dousing System and the Open Top installation method and advantages relative to traditional CANDU 6 construction practices are also presented. Finally, major improvements for future CANDU plant construction using the Open Top method are discussed. (author)

'Thin walled' concept and a new top lid applied to the Scandinavian PCRV for a boiling water reactor

International Nuclear Information System (INIS)

Scotto, F.L.

1975-01-01

This research is carried out in the frame of an agreement between AB ATOMENERGI of Sweden and ENEL (Ente Nazionale per l'Energia Elettrica) of Italy, for an exchange of information in the field of PCPV for BWR, and takes as a reference the Scandinavian solution as far as the thermal insulation system and the geometry are concerned, proposing new solutions for the prestressed concrete structure (namely the Author's concept of thin walls and a new concept of top lid). The proposed top lid sealing system solution is in line with the one adopted for the conventional steel pressure vessel enclosures; furthermore the prestressed concrete lid is restricted to the prestressed concrete structure to form a continuous contrete structure, in line with th PCPV conventional solutions for gas reactors. The paper describes in detail the selected design philosophy that is slightly different from the one defined by the Scandinavian project. In fact, as far as the design limits are concerned, it refers mainly to steel pressure vessel philosophy and, as to the concrete behaviour, to the design philosophy proposed by the author for the PCPV 'thin walled' structures for gas-cooled power reactors. Rheological, mathematical and physical models had been suitably devised in order to check the reliability of the proposed assumption. This paper therefore, will also give a brief description of said tools and the main results acquired at the time of the conference, and technical and economical considerations made to support the interest of the research, showing the relevant cut down of the costs. The comparative reference steel pressure vessel belongs to Mark III ENEL VI and VIII BW plant to which design and construction the author gives his contribution

Top-down and bottom-up approaches for cost estimating new reactor designs

International Nuclear Information System (INIS)

Berbey, P.; Gautier, G.M.; Duflo, D.; Rouyer, J.L.

2007-01-01

For several years, Generation-4 designs will be 'pre-conceptual' for the less mature concepts and 'preliminary' for the more mature concepts. In this situation, appropriate data for some of the plant systems may be lacking to develop a bottom-up cost estimate. Therefore, a more global approach, the Top-Down Approach (TDA), is needed to help the designers and decision makers in comparing design options. It utilizes more or less simple models for cost estimating the different parts of a design. TDA cost estimating effort applies to a whole functional element whose cost is approached by similar estimations coming from existing data, ratios and models, for a given range of variation of parameters. Modeling is used when direct analogy is not possible. There are two types of models, global and specific ones. Global models are applied to cost modules related to Code Of Account. Exponential formulae such as Ci = Ai + (Bi x Pi n ) are used when there are cost data for comparable modules in nuclear or other industries. Specific cost models are developed for major specific components of the plant: - process equipment such as reactor vessel, steam generators or large heat exchangers. - buildings, with formulae estimating the construction cost from base cost of m3 of building volume. - systems, when unit costs, cost ratios and models are used, depending on the level of detail of the design. Bottom Up Approach (BUA), which is based on unit prices coming from similar equipment or from manufacturer consulting, is very valuable and gives better cost estimations than TDA when it can be applied, that is at a rather late stage of the design. Both approaches are complementary when some parts of the design are detailed enough to be estimated by BUA, and when BUA results are used to check TDA results and to improve TDA models. This methodology is applied to the HTR (High Temperature Reactor) concept and to an advanced PWR design

Calorimetric measurements of the dose absorbed to graphite in the research reactors Melusine (8MWth) and Siloe (35MWth), and in the power reactor Bugey 1 (1900 MWth)

International Nuclear Information System (INIS)

Petitcolas, H.; Bonnin, J.J.; Chenavas, P.

1975-01-01

The TM calorimeter (Melusine type) developed in the CEN/Grenoble research reactors allows measurement of dose-rates over the range 10 -3 to 10 W/g. Simple and of small volume, the calorimeter causes minimum perturbation of the radiation field in which the measurement is to be made; it adapts easily to the specific requirements of the irradiation loops for which it is primarily designed. The operation is simple: measurement of temperature difference and time constants at various equilibrium positions. The results compare very favourably with those obtained by other methods, (eg. ionisation chambers) and by other workers. The calorimeter will operate for several years under irradiation and under the severe conditions of pressure, temperature etc., occurring in certain power reactors [fr

Proceedings of TopSafe 2008 Transactions

International Nuclear Information System (INIS)

2008-01-01

The aim of the conference is to provide a forum for addressing the current status and future perspectives with regards to safety at nuclear installations worldwide. Previous TopSafe editions took place in Budapest (1995) and Valencia (1998). The conference is directed at a broad range of experts in the area of nuclear safety, including professionals from the different disciplines involved in the safety of nuclear power plants, installations in other parts of the fuel cycle, and research reactors. It is aimed at professionals coming from the research organisations, universities, vendors, operators, regulatory bodies as well as policy makers. Top level representatives of the Countries that are constructing new nuclear power plants are invited. Regulators of all individual Countries with nuclear programme are expected to contribute the Conference. The topics of the conference are: Safety Issues of Operating Power Plants PWR and BWR, CANDU, WWER, RBMK; Application of European Utilities Requirements; Probabilistic and Deterministic Analysis; Shutdown Safety; Advances in Safety: Analysis Codes and Techniques; Severe Accidents Management; International Safety Studies; Emergency Planning; Risk Informed Application and Licensing; Regulatory Safety Requirements; Ageing and Life Extension; Power Upgrades and Relevant Topics; Management of Safety and Quality; Safety Culture and Self Assessment; Political and Public Perception of Nuclear Energy; Nuclear Power Plant Security; Safety Issues of Future Power Plants-Near term deployment reactors (EPR, SWR1000, AP1000, ESBWR, SBWR, ACR-1000) and Generation IV reactors; Safety Issues of Research Reactors (pool type and others); Fuel Cycle Facilities Safety-Uranium mining and conversion, enrichment and fuel production, reprocessing and transmutation, waste disposal. (authors)

TRIGA reactor health physics considerations

International Nuclear Information System (INIS)

Johnson, A.G.

1970-01-01

The factors influencing the complexity of a TRIGA health physics program are discussed in details in order to serve as a basis for later consideration of various specific aspects of a typical TRIGA health physics program. The health physics program must be able to provide adequate assistance, control, and safety for individuals ranging from the inexperienced student to the experienced postgraduate researcher. Some of the major aspects discussed are: effluent release and control; reactor area air monitoring; area monitoring; adjacent facilities monitoring; portable instrumentation, personnel monitoring. TRIGA reactors have not been associated with many significant occurrences in the area of health physics, although some operational occurrences have had health physics implications. One specific occurrence at OSU is described involving the detection of non-fission-product radioactive particulates by the continuous air monitor on the reactor top. The studies of this particular situation indicate that most of the particulate activity is coming from the rotating rack and exhausting to the reactor top through the rotating rack loading tube

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Evaluation inquiry of the provision of iodine tablets to the populations living in the vicinity of the nuclear center of electric production at Bugey, september 1998

International Nuclear Information System (INIS)

Helynck, B.; Rey, S.; Malfait, Ph.; Dubois, M.C.

2002-01-01

50 % of households in the area of the particular plan of intervention (P.P.I.) of Bugey N.P.P. had removed the iodine tablets. This cover is insufficient, when the Ministry circular considered an availability of tablets in every household. The inquiry highlighted a relationship with between the removal of tablets and knowledge about iodine, but the results suggest that this information campaign was not enough efficient. This situation is worrying, because of the expiry date of the tablets, new campaigns are to be done regularly. In order to improve the results of future campaigns, it is important to take into account feedbacks of the first distribution. The information sources must be diversified. Health professionals should be involved. The Institutions in charge of health in children (Service of maternal and infants protection and service for the promotion of health for students) should also be involved. Simplification and clarity of messages must be improved, especially towards priority populations such young parents, pregnant women. (N.C.)

Nuclear reactor

International Nuclear Information System (INIS)

Rau, P.

1980-01-01

The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

Physical aspects of liquid-impelled loop reactors

NARCIS (Netherlands)

Sonsbeek, van H.

1992-01-01

The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the

Safety of RBMK reactors: Major results and prospects

International Nuclear Information System (INIS)

Sidorenko, V.A.

1996-01-01

The paper considers the following issues: basic reasons for the advent of NPPs with RBMK reactors; the logic of identifying top-priority measures immediately after the accident; top-priority measures for improving the safety and reliability of NPPs with RBMK reactors; upgrading NPPs with RBMK reactors in compliance with the Norms; programmes for retrofitting and upgrading of NPPs of the ''Rosnergoatom'' Concern and progress with their implementation as of April 1996; the safety of RBMK plants and the programmes of its enhancement with regard to modern requirements in the light of national and international assessment; objective indicators of safety, reliability, and economic efficiency of NPPs with RBMK reactors; economics: rationale for continuing plants operation till the end of their design lifetime. 8 refs, 3 figs

Pressurized water reactor flow arrangement

International Nuclear Information System (INIS)

Gibbons, J.F.; Knapp, R.W.

1980-01-01

A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

Fuel handling system of nuclear reactor plants

International Nuclear Information System (INIS)

Faulstich, D.L.

1991-01-01

This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

Reactor building for a nuclear reactor

International Nuclear Information System (INIS)

Haidlen, F.

1976-01-01

The invention concerns the improvement of the design of a liner, supported by a latticed steel girder structure and destined for guaranteeing a gastight closure for the plant compartments in the reactor building of a pressurized water reactor. It is intended to provide the steel girder structure on their top side with grates, being suited for walking upon, and to hang on their lower side diaphragms in modular construction as a liner. At the edges they may be sealed with bellows in order to avoid thermal stresses. The steel girder structure may at the same time serve as supports for parts of the steam pipe. (RW) [de

The bottom-supported fast reactor - system simplifications and enhanced safety

International Nuclear Information System (INIS)

Petrozelli, J.; Golan, S.; Kawamura, Yutaka; Kumaoka, Yoshio; Nakagawa, Hiroshi

1992-01-01

The 600-MW(electric) bottom-supported fast reactor (BSFR) incorporates the following key features: (1) modular upper internal structure (UIS); (2) electromagnetic pumps (EMPs); (3) low-sodium-void-worth metal-fuel core; and (4) bottom supported reactor vessel (BSRV), which is entirely supported by the basement, except for the control rods, control rod drives (CRDs), UIS, and the stationary plug; by comparison, a top-supported reactor vessel (TSRV) is completely supported by the operating floor. The diameter of the reactor vessel (RV) is 12.8 m (42 ft), and the height (distance from the basemat to the operating floor) is 19.8 m (65 ft). The RV is supported by a single support cylinder anchored to the basemat. The core has 210 driver assemblies and 192 radial blanket assemblies in an annular configuration. The primary heat transport system components consist of four intermediate heat exchangers (IHXs), four EMPs, and four primary reactor auxillary cooling systems. All these components are supported by the BSRV and hang from their tops. Six modular, vertically movable UIS mechanisms clear the UIS from the space over the core during refueling. The top closure is designed to operate at the reactor outlet temperature and is free to expand and contract. Small bellows between the top closure and each UIS model accommodate differential movements and comprise a portion of the cover gas boundary. A 1200-MW(electric) plant with two 600-MW(electric) (twin) nuclear steam supply systems is being studied

Top-nozzle mounted replacement guide pin assemblies

International Nuclear Information System (INIS)

Gilmore, C.B.; Andrews, W.H.

1993-01-01

A replacement guide pin assembly is provided for aligning a nuclear fuel assembly with an upper core plate of a nuclear reactor core. The guide pin assembly includes a guide pin body having a radially expandable base insertable within a hole in the top nozzle, a ferrule insertable within the guide pin base and capable of imparting a radially and outwardly directed force on the expandable base to expand it within the hole of the top nozzle and thereby secure the guide pin body to the top nozzle in response to a predetermined displacement of the ferrule relative to the guide pin body along its longitudinal axis, and a lock screw interfitted with the ferrule and threaded into the guide pin body so as to produce the predetermined displacement of the ferrule. (author)

Top-MOX fuel solution: strategies, challenges, opportunities

International Nuclear Information System (INIS)

Breitenstein, P.; Vo Van, V.

2014-01-01

TOP-MOX is a nuclear fuel solution and product developed by AREVA and successfully implemented in Europe. It allows utilities burning plutonium (instead of enriched uranium) even when this plutonium is not stemming from own reprocessed used fuel - that is third party plutonium. The important challenges for utilities along with TOP-MOX implementation are legal/patrimonial Pu-ownership issues and general economical aspects. Available sponsorship of such plutonium permits UO2 competitive market prices. For new MOX customers licensing and technical aspects come along. Further AREVA proposes a flexible solution which is called 'TOP-MOX pre-cycling'. This involves making available third party plutonium for fuel fabrication and reactor use pending the utilities' final strategic fuel cycle decision. The paper gives insight into and analyses the impacts of allowing customers the implementation of a TOP-MOX program with focus on Pu-ownership, economics, technical and legal aspects as well as the impact on used MOX management and final waste management. (authors)

Assessment of the radiological inventory of EDF's graphite waste through an assimilation method

International Nuclear Information System (INIS)

Poncet, B.

2014-01-01

The definitive disposal of graphite from the decommissioned UNGG reactors (Chinon A3, Saint-Laurent A1, Saint-Laurent A2 and Bugey 1) has required a radiological inventory of the irradiated graphite. This study focuses on Cl 36 that is produced by neutron absorption on Cl 35 that was present initially in graphite as an impurity (about 80 mg/t of Cl initially in Bugey 1 graphite)). It appears that the changes of Cl 36 concentration along the height of a stack of graphite do neither fit the changes in the neutron flux nor the changes in the graphite temperature. This fact is explained by the high level of purity of the graphite and the nugget effect. Challenged by the absence of spatial correlation of the Cl 36 concentration, an EDF's team has developed an assimilation method based on comparisons between calculations and measurements in order to get a conservative inventory. (A.C.)

Zinc injection on the EDF fleet monitoring the injection on 12 units

International Nuclear Information System (INIS)

Le-Meur, Gaelle Harmand; Anne-Marie; Stutzmann, Agnes; Taunier, Stephane; Benfarah, Moez; Bretelle, Jean-Luc; Alain, Rocher; Claeys, Myriam; Bonne, Sebastien

2012-09-01

After a first implementation of zinc injection at Bugey 2 and Bugey 4, EDF decided to extend the program to other units of its fleet. 14 more reactors from the French fleet of 58 were chosen in order to - Reduce the radiation sources for curative or preventive (after SGR) reasons - Mitigate stress corrosion cracking on nickel alloys and reduce the rate of generalized corrosion - Prevent the risk of CIPS, mainly after a fuel management change. Zinc injection started on 9 new units in 2011, 1 unit in 2012 and will be extended to 4 other units before the end of 2013. To monitor the injection, EDF has defined a complete program concerning chemistry, radiation protection (dose rate and deposited activities measurements), materials (statistical analysis of SG tube cracks), fuel (oxide measurements) and waste (radiochemical characterization of filters). Reference units were chosen for each field because of the size of the fleet. This paper will detail the different monitoring programs on the EDF plants injecting zinc. (authors)

Nuclear reactor container

International Nuclear Information System (INIS)

Moriyama, Takeo; Ochiai, Kanehiro; Niino, Tsuyoshi; Kodama, Toyokazu; Hirako, Shizuka.

1988-01-01

Purpose: To obtain structures suitable to a container structures for nuclear power plants used in those districts where earthquakes occur frequently, in which no local stresses are caused to the fundamental base portions and the workability for the fundamental structures is improved. Constitution: Basic stabilizers are attached to a nuclear reactor container (PCV) and a basic concrete recess for receiving a basic stabilizer is disposed in basic concretes. A top stabilizer is joined and fixed to a top stabilizer receiving plate at the inside of an outer shielding wall. On the other hand, a PCV top recess for conducting the load of PCV to the top stabilizer is attached to the top of the PCV. By disposing stabilizer structures allowing miner displacements at the two points, that is, the top and the lowermost portion of the PCV, no local stress concentrations can be generated to the extension on the axial direction of components due to the inner pressure of the PCV and to the horizontal load applied to the upper portion of the PCV upon earthquakes. (Yoshino, Y.)

Fuel handling grapple for nuclear reactor plants

International Nuclear Information System (INIS)

Rousar, D.L.

1992-01-01

This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

Replacement of a vessel head, an operation which today gets easily into its stride

International Nuclear Information System (INIS)

Mardon, P.; Chaumont, J.C.; Lambiotte, P.

1995-01-01

In 1992, one year after the detection of a leak in a vessel head of the Electricite de France (EDF) Bugey 4 reactor, the head was replaced by the Framatome-Jeumont Industrie Group. Today, this group, which has developed new methods and new tools to optimize the cost, the time-delay and the dosimetry of this kind of intervention, has performed 11 additional replacements, two of which on 1300 MWe power units. This paper describes step by step the successive operations required for a complete vessel head replacement, including the testing of safety systems before starting up the reactor. (J.S.). 7 photos

Safety of nuclear power reactors

International Nuclear Information System (INIS)

MacPherson, H.G.

1982-01-01

Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

Review of nuclear reactor accidents

International Nuclear Information System (INIS)

Connelly, J.W.; Storr, G.J.

1989-01-01

Two types of severe reactor accidents - loss of coolant or coolant flow and transient overpower (TOP) accidents - are described and compared. Accidents in research reactors are discussed. The 1961 SL1 accident in the US is used as an illustration as it incorporates the three features usually combined in a severe accident - a design flaw or flaws in the system, a circumvention of safety circuits or procedures, and gross operator error. The SL1 reactor, the reactivity accident and the following fuel-coolant interaction and steam explosion are reviewed. 3 figs

Reactor containing facility

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1992-01-01

A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

Pulsed fusion reactors

International Nuclear Information System (INIS)

1975-01-01

This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

Device for supporting a nuclear reactor core

International Nuclear Information System (INIS)

Costes, D.

1976-01-01

The core of a light-water reactor which is enclosed in a prestressed concrete pressure vessel and held within a diffuser basket is supported by a device consisting of a cylindrical shell which surrounds the basket and is rigidly fixed to a plurality of frusto-conical skirts having concurrent axes and located substantially at right angles to the axis of the reactor core. The small base of each skirt is rigidly fixed to the shell and the large base is anchored in openings formed in the reactor vessel for the penetration of coolant inlet and outlet pipes. The top portion of the shell is secured to the top portion of the diffuser basket, a flat surface being formed on the shell at the point of connection with each frusto-conical skirt so as to ensure rigid suspension while permitting thermal expansion

Pre evaluation for heat balance of prototype sodium cooled fast reactor

International Nuclear Information System (INIS)

Han, Ji Woong; Kim, De Hee; Yoon, Jung; Kim, Eui Kwang; Lee, Tae Ho

2012-01-01

Under the long term advanced SFR R and D plan, the design of prototype reactor has been carried out toward the construction of the prototype SFR plant by 2028. The R and D efforts in fluid system design will be focused on developing a prototype design of primary heat transport system(PHTS), intermediate heat transport system (IHTS), decay heat removal system(DHRS), steam generation system(SGS), and related auxiliary system design for a prototype reactor as shown in Fig. 1. In order to make progress system design, top tier requirements for prototype reactor related to design parameters of NSSS and BOP should be decided at first. The top tier requirement includes general design basis, capacity and characteristics of reactor, various requirements related to safety, performance, securities, economics, site, and etc.. Extensive discussion has been done within Korea Atomic Energy Research Institute(KAERI) for the decision of top tier requirements of the prototype reactor. The core outlet temperature, which should be described as top tier requirements, is one of the critical parameter for system design. The higher core exit temperature could contribute to increase the plant efficiency. However, it could also contribute to decrease the design margin for structure and safety. Therefore various operating strategies based on different core outlet temperatures should be examined and evaluated. For the prototype reactor two core outlet temperatures are taken into accounted. The lower temperature is for the operation condition and the higher temperature is for the system design and licensing process of the prototype reactor. In order to evaluate the operability of prototype reactor designed based on higher temperature, the heat balance calculations have been performed at different core outlet temperature conditions. The electrical power of prototype reactor was assumed to be 100MWe and reference operating conditions were decided based on existing available data. The

Summarized presentation of the numerical model used for the pressurizer of a light water nuclear reactor. Description and validation

International Nuclear Information System (INIS)

Siarry, P.

1981-12-01

The pressurizer model is first described together with its coupling to the nuclear unit. The different stages involved in the validation are then presented: validation of overall qualitative behavior; validation of the open loop pressurizer model; validation of the various units for controlling pressures and levels; simulation of two large transients (Bugey plant) [fr

Nuclear reactor vessel decontamination systems

International Nuclear Information System (INIS)

McGuire, P. J.

1985-01-01

There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

Reconstitutable nuclear reactor fuel assembly with unitary removable top nozzle subassembly

International Nuclear Information System (INIS)

Shallenberger, J.M.

1987-01-01

A reconstitutable fuel assembly is described having at least one control rod guide thimble and a top nozzle, the guide thimble including an upper extension, the top nozzle including at least one hold-down spring, an upper hold-down plate and a lower adapter plate, an improved attaching structure removably mounting the top nozzle as a unitary subassembly on the guide thimble. The attaching structure comprises: (a) a coupling member interfitting the lower adapter plate, the upper hold-down plate and the hold-down spring disposed between the plates so as to capture and retain the plates and spring together as a unitary subassembly in which the upper plate is slidably moveable along the coupling member relative to the lower plate with the spring biasing the upper plate away from the lower plate. The coupling member has spaced apart upper and lower portions with a central passageway extending for slidably receiving the upper extension of the guide thimble in a nonattached relationship in which the coupling member is slidably movable relative to the guide thimble extension for respectively inserting and removing the coupling member on and from the guide thimble extension

Method of dismantling a nuclear reactor

International Nuclear Information System (INIS)

Shirai, Masato; Hashimoto, Osamu.

1984-01-01

Purpose: To enable rapid and simple positioning for a plasma arc torch disposed to the inside of a nuclear reactor main body. Method: After removing the upper semi-spherical portion, fuel portion and control rod portion of a nuclear reactor, a rotary type girder is placed on the upper edge of a cylindrical portion remained after the removal of the upper semi-spherical portion. Then, the upper portion of a supporting rod provided with a swing arm having a plasma arc torch at the top end is situated at the center of the reactor main body. Then, the top end of the support rod is inserted to fix in the housing of control rod drives. Then, the swing arm is actuated to situate the plasma arc torch to a desired position to be cut, whereafter cutting is initiated while rotating the rotary type girder. Thus, plasma arc torch is moved horizontally along an arcuate trace, whereby pipeways, accessories or the likes disposed to the inside of the main body are at first cut and then the cylindrical portion constituting the main body is cut to dismantle the reactor. (Moriyama, K.)

Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo [Korea Advanced Institute of Science and Tehcnology, Daejeon (Korea, Republic of)

2006-03-15

There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis.

Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

International Nuclear Information System (INIS)

Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo

2006-03-01

There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis

Fluiddynamic effects in the fuel element top nozzle area during refilling and reflooding

International Nuclear Information System (INIS)

Hawighorst, A.; Kroening, H.; Mewes, D.; Spatz, R.; Mayinger, F.

1985-01-01

During the refilling and reflooding phase following a hypothetical loss of coolant accident in lightwater cooled nuclear reactors, there will be countercurrent flow between discharging steam and the feed of emergency core cooling water. It was the objective of this research project to contribute to a better physical understanding of the fluiddynamic processes in the area of the fuel element top nozzle and so to improve emergency core cooling calculations. Therefore, experimental and theoretical investigations about the entrainment and countercurrent behaviour of gas/liquid flows have been implemented within this project. Fluiddynamic processes in the fuel element top nozzle area were simulated during the reflooding and refilling phase. Based on special internals as single and multiple-hole orifices, basic phenomena of fluidynamics were studied first with air-water. Subsequently, investigations of the system steam/water were conducted. The reactor geometry was approximated step by step, until a complete reactor fuel assembly top nozzle was constituted. The system pressure was 4.8 bars (abs), in accordance with the conditions in the reactor pressure vessel at the end of the blowdown phase. The water was initially fed in at saturation temperature, then, as a second step, fed in at subcooled condition relative to the steam temperature, in order to be able to study condensation effects as well. First, investigations on gas/liquid countercurrent flows in the fluid system air/water are presented. Then one studies countercurrent flow in the system steam/water, including the investigation of condensation effects. Finally, a detailed description of the research on droplet size determination is given

Project and feedback experience on nuclear facility decommissioning

Energy Technology Data Exchange (ETDEWEB)

Santiago, J.L. [ENRESA (Spain); Benest, T.G. [United Kingdom Atomic Energy Authority, Windscale, Cumbria (United Kingdom); Tardy, F.; Lefevre, Ph. [Electricite de France (EDF/CIDEN), 69 - Villeurbanne (France); Willis, A. [VT Nuclear Services (United Kingdom); Gilis, R.; Lewandowski, P.; Ooms, B.; Reusen, N.; Van Laer, W.; Walthery, R. [Belgoprocess (Belgium); Jeanjacques, M. [CEA Saclay, 91 - Gif sur Yvette (France); Bohar, M.P.; Bremond, M.P.; Poyau, C.; Mandard, L.; Boissonneau, J.F.; Fouquereau, A.; Pichereau, E.; Binet, C. [CEA Fontenay aux Roses, 92 (France); Fontana, Ph.; Fraize, G. [CEA Marcoule 30 (France); Seurat, Ph. [AREVA NC, 75 - Paris (France); Chesnokov, A.V.; Fadin, S.Y.; Ivanov, O.P.; Kolyadin, V.I.; Lemus, A.V.; Pavlenko, V.I.; Semenov, S.G.; Shisha, A.D.; Volkov, V.G.; Zverkov, Y.A. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

2008-11-15

This series of 6 short articles presents the feedback experience that has been drawn from various nuclear facility dismantling and presents 3 decommissioning projects: first, the WAGR project that is the UK demonstration project for power reactor decommissioning (a review of the tools used to dismantle the reactor core); secondly, the dismantling project of the Bugey-1 UNGG reactor for which the dismantling works of the reactor internals is planned to be done underwater; and thirdly, the decommissioning project of the MR reactor in the Kurchatov Institute. The feedback experience described concerns nuclear facilities in Spain (Vandellos-1 and the CIEMAT research center), in Belgium (the Eurochemic reprocessing plant), and in France (the decommissioning of nuclear premises inside the Fontenay-aux-roses Cea center and the decommissioning of the UP1 spent fuel reprocessing plant at the Marcoule site). (A.C.)

Project and feedback experience on nuclear facility decommissioning

International Nuclear Information System (INIS)

Santiago, J.L.; Benest, T.G.; Tardy, F.; Lefevre, Ph.; Willis, A.; Gilis, R.; Lewandowski, P.; Ooms, B.; Reusen, N.; Van Laer, W.; Walthery, R.; Jeanjacques, M.; Bohar, M.P.; Bremond, M.P.; Poyau, C.; Mandard, L.; Boissonneau, J.F.; Fouquereau, A.; Pichereau, E.; Binet, C.; Fontana, Ph.; Fraize, G.; Seurat, Ph.; Chesnokov, A.V.; Fadin, S.Y.; Ivanov, O.P.; Kolyadin, V.I.; Lemus, A.V.; Pavlenko, V.I.; Semenov, S.G.; Shisha, A.D.; Volkov, V.G.; Zverkov, Y.A.

2008-01-01

This series of 6 short articles presents the feedback experience that has been drawn from various nuclear facility dismantling and presents 3 decommissioning projects: first, the WAGR project that is the UK demonstration project for power reactor decommissioning (a review of the tools used to dismantle the reactor core); secondly, the dismantling project of the Bugey-1 UNGG reactor for which the dismantling works of the reactor internals is planned to be done underwater; and thirdly, the decommissioning project of the MR reactor in the Kurchatov Institute. The feedback experience described concerns nuclear facilities in Spain (Vandellos-1 and the CIEMAT research center), in Belgium (the Eurochemic reprocessing plant), and in France (the decommissioning of nuclear premises inside the Fontenay-aux-roses Cea center and the decommissioning of the UP1 spent fuel reprocessing plant at the Marcoule site). (A.C.)

In-pile inspections of the Calder and Chapelcross nuclear reactors

International Nuclear Information System (INIS)

Stewart, G.

1984-01-01

The subject is discussed under the headings: introduction (relevant data about the reactors); inspection policy; photographic inspection (equipment; inspection results (vessel seam welds and plates; top dome welds; top dome internals)); ultrasonic equipment; manipulator; television inspections; concluding remarks. (U.K.)

Method and apparatus for removably mounting a top nozzle on a nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Wilson, J.F.; Gjertsen, R.K.; Schallenberger, J.M.

1986-01-01

In a fuel assembly having a top nozzle and control rod guide thimbles, a method is described of removably mounting the top nozzle on the ends of the guide thimbles, comprising the steps of: (a) releasably mating hollow outer sockets defined in the top nozzle with hollow inner sockets defined on the ends of the guide thimbles. The inner sockets are movable between compressed conditions for removing and inserting the inner sockets from and into the outer sockets in mounting and removing the top nozzle on and from the guide thimbles and expanded conditions for mating the inner and outer sockets together and the top nozzle on the guide thimbles; (b) supporting elongated locking tubes such that end portions thereof extend into the outer sockets defined in the top nozzle; and (c) moving all of the locking tubes at the same time between unlocking and locking positions to displace their end portions axially within the outer sockets between first and second locations

Development of observation techniques in reactor vessel of experimental fast reactor Joyo

International Nuclear Information System (INIS)

Takamatsu, Misao; Imaizumi, Kazuyuki; Nagai, Akinori; Sekine, Takashi; Maeda, Yukimoto

2010-01-01

In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. And several IVO equipments for an SFR are developed. However, in order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident that occurred with Joyo, IVO using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1 mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8 mm in thickness could be discriminated. The performance of IVO equipments under the actual reactor environment was successfully confirmed. And the results provided useful information on incident investigations. In addition, fundamental findings and the experience gained during this study, which included the design of equipment, operating procedures, resolution, lighting adjustments, photograph composition and the durability of the RRF under radiation exposure, provided valuable insights into further improvements and verifications for IVO techniques to

Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

International Nuclear Information System (INIS)

Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

2007-01-01

Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

Graphite surveillance in N Reactor

International Nuclear Information System (INIS)

Woodruff, E.M.

1991-09-01

Graphite dimensional changes in N Reactor during its 24 yr operating history are reviewed. Test irradiation results, block measurements, stack profiles, top of reflector motion monitors, and visual observations of distortion are described. 18 refs., 14 figs., 1 tab

Desk top calculation strategies for reactor analysis using Mathematica

International Nuclear Information System (INIS)

Kullberg, C.

1991-01-01

Mathematics is one of several recently developed equations analysis programs that is particularly well suited for solving a broad range of intermediate engineering problems. The objective of this paper is to demonstrate, using a couple of reactor-related examples, how Mathematica can be exploited as a user-friendly analysis tool to symbolically and numerically handle systems of algebraic and differential equations. 7 refs., 3 figs

Neutronic reactor thermal shield

International Nuclear Information System (INIS)

Lowe, P.E.

1976-01-01

A shield for a nuclear reactor includes at least two layers of alternating wide and narrow rectangular blocks so arranged that the spaces between blocks in adjacent layers are out of registry, each block having an opening therein equally spaced from the sides of the blocks and nearer the top of the block than the bottom, the distance from the top of the block to the opening in one layer being different from this distance in adjacent layers, openings in blocks in adjacent layers being in registry. 1 claim, 7 drawing figures

Fusion reactor pumped laser

International Nuclear Information System (INIS)

Jassby, D.L.

1988-01-01

A nuclear pumped laser is described comprising: a toroidal fusion reactor, the reactor generating energetic neutrons; an annular gas cell disposed around the outer periphery of the reactor, the cell including an annular reflecting mirror disposed at the bottom of the cell and an annular output window disposed at the top of the cell; a gas lasing medium disposed within the annular cell for generating output laser radiation; neutron reflector material means disposed around the annular cell for reflecting neutrons incident thereon back into the gas cell; neutron moderator material means disposed between the reactor and the gas cell and between the gas cell and the neutron reflector material for moderating the energy of energetic neutrons from the reactor; converting means for converting energy from the moderated neutrons to energy pumping means for pumping the gas lasing medium; and beam compactor means for receiving output laser radiation from the annular output window and generating a single output laser beam therefrom

Reactor core cooling device

International Nuclear Information System (INIS)

Kobayashi, Masahiro.

1986-01-01

Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

Composting on Mars or the Moon: II. Temperature feedback control with top-wise introduction of waste material and air

Science.gov (United States)

Finstein, M. S.; Hogan, J. A.; Sager, J. C.; Cowan, R. M.; Strom, P. F.; Janes, H. W. (Principal Investigator)

1999-01-01

Whereas Earth-based composting reactors that effectively control the process are batch operations with bottom-to-top airflow, in extraterrestrial application both the fresh waste and the air need to be introduced from above. Stabilized compost and used air would exit below. This materials flow pattern permits the addition of waste whenever generated, obviating the need for multiple reactors, and the incorporation of a commode in the lid. Top loading in turn dictates top-down aeration, so that the most actively decomposing material (greatest need for heat removal and O2 replenishment) is first encountered. This novel material and aeration pattern was tested in conjunction with temperature feedback process control. Reactor characteristics were: working, volume, 0.15 m3; charge, 2 kg dry biomass per day (comparable to a 3-4 person self-sufficient bioregenerative habitat); retention time, 7 days. Judging from temperature profile, O2 level, air usage, pressure head loss, moisture, and odor, the system was effectively controlled over a 35-day period. Dry matter disappearance averaged 25% (10-42%). The compost product was substantially, though not completely, stabilized. This demonstrates the compatibility of top-wise introduction of waste and air with temperature feedback process control.

EDF decommissioning programme a global commitment to safety, environment and cost efficiency of nuclear energy

International Nuclear Information System (INIS)

Grenouillet, J.-J.

2002-01-01

EDF has 9 NPPs permanently shutdown and under decommissioning. EDF considers that if the nuclear option is to remain open, it is necessary to deal with increasing public concerns for environmental and waste management issues. Therefore EDF has decided to achieve total dismantling of all shutdown reactor in the next 25 years. The Decommissioning Program has been developed including 2 stages of activities. The first stage consists of: 1) Final dismantling of Brennilis in 2015; 2) A dismantling demonstration of a PWR reactor building (Chooz A) before starting replacing the population of PWRs currently in operation; 3) Final dismantling of reactor containment of a GCR (Bugey 1) as a first of its kind. The second stage includes: 1)Dismantling of following 5 GCR (Saint Laurent A1 and A2, Chinon A1, A2 and A3); 2) Final dismantling of Chooz A and Bugey 1 in 2025. The successful implementation relies on the simplification of the regulatory process; availability of treatment, conditioning and disposal facilities and effective nuclear industry. The main issue is availability of time and waste solutions such as opening of a Very Low Waste disposal in 2003 (130 000 tons); opening of a new disposal for graphite and radiferous wastes (17 000 tons) in 2010 and opening in 2007-2008 of a centralized interim storage (BANEDA) facility for long-lived Medium Level Wastes (500 tons including filters, control rods etc)Three investigations are to be carried out for high level radioactive waste before 2006

Lining up device for the internal structures of a nuclear reactor vessel

International Nuclear Information System (INIS)

Silverblatt, B.L.

1977-01-01

The invention concerns a nuclear reactor of the type with a vessel, a vessel head carried at the top of this vessel by a core cylinder comprising a flange internally supported by the vessel, and an upper support structure supported between the core cylinder flange and the vessel head to align laterally the head, vessel, flange and support structure. A bottom key device is provided for lining up the flange, support structure and vessel, and an upper key device for laterally lining up support structure and the vessel head and for maintaining this alignment when they are removed simultaneously from the core cylinder and vessel. When re-assembling the reactor, the top support structure and the vessel head are lowered simultaneously so that an opening in the top alignment structure engages in the upper extension of the bottom alignment structure. A plurality of alignment stuctures may be utilised round the circumference of the reactor vessel. The disposition of the invention also facilitates the removal of the core cylinder from the reactor vessel. In this way, the alignment on re-assembly is ensured by the re-entry of the bottom extension under the flange of the core cylinder with the groove or keyway of the reactor vessel [fr

Photomultipliers gain monitoring at the one percent level with a blue light pulser

International Nuclear Information System (INIS)

Berger, J.; Bermond, M.; Besson, P.; Favier, J.; Pessard, H.; Poulet, M.

1988-07-01

We describe a method and an experimental layout allowing the monitoring of photomultipliers gain. We use artificial blue light (Spark-gap with filter: 436 ± 20 nm) and three reference detectors. Short term and long term measurements are presented. The results indicate a precision better than 0.5% for the short term and 1.4% for the long term determinations. This gain monitoring system has been developed for a new neutrino oscillation reactor experiment (600 photomultipliers) starting at the Bugey nuclear plant

French statutory approach of the evaluation of the safety level of old nuclear divisions; Approche reglementaire francaise de l`appreciation du niveau de surete des tranches anciennes

Energy Technology Data Exchange (ETDEWEB)

Delage, M

1994-06-01

The legal French procedures include three steps which have to be followed during the exam of the safety in nuclear plants (creation authorization, loading authorization, actual running of the plant). After listing the different types of evaluation of safety in fraction of plants, this report presents the main themes encountered during the safety assessment: state of the reactor, maintenance, tracking of the incidents, personnel training, radioprotection, radioactive releases. The Fessenheim and Bugey list of reevaluation themes is also given. (TEC).

Experience with reactor assembly of FBTR

International Nuclear Information System (INIS)

Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

2006-01-01

Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

Nitrogen removal in the bioreactor landfill system with intermittent aeration at the top of landfilled waste

International Nuclear Information System (INIS)

He Ruo; Shen Dongsheng

2006-01-01

High ammonia concentration of recycled landfill leachate makes it very difficult to treat. In this work, a vertical aerobic/anoxic/anaerobic lab-scale bioreactor landfill system, which was constructed by intermittent aeration at the top of landfilled waste, as a bioreactor for in situ nitrogen removal was investigated during waste stabilization. Intermittent aeration at the top of landfilled waste might stimulate the growth of nitrifying bacteria and denitrifying bacteria in the top and middle layers of waste. The nitrifying bacteria population for the landfill bioreactor with intermittent aeration system reached between10 6 and 10 8 cells/dry g waste, although it decreased 2 orders of magnitude on day 30, due to the inhibitory effect of the acid environment and high organic matter in the landfilled waste. The denitrifying bacteria population increased by between 4 and 13 orders of magnitude compared with conventional anaerobic landfilled waste layers. Leachate NO 3 - -N concentration was very low in both two experimental landfill reactors. After 105 days operation, leachate NH 4 + -N and TN concentrations for the landfill reactor with intermittent aeration system dropped to 186 and 289 mg/l, respectively, while they were still kept above 1000 mg/l for the landfill reactor without intermittent aerobic system. In addition, there is an increase in the rate of waste stabilization as well as an increase of 12% in the total waste settlement for the landfill reactor with intermittent aeration system

Experimental validation of thermal design of top shield for a pool type SFR

International Nuclear Information System (INIS)

Aithal, Sriramachandra; Babu, V. Rajan; Balasubramaniyan, V.; Velusamy, K.; Chellapandi, P.

2016-01-01

Highlights: • Overall thermal design of top shield in a SFR is experimentally verified. • Air jet cooling is effective in ensuring the temperatures limits for top shield. • Convection patterns in narrow annulus are in line with published CFD results. • Wire mesh insulation ensures gradual thermal gradient at top portion of main vessel. • Under loss of cooling scenario, sufficient time is available for corrective action. - Abstract: An Integrated Top Shield Test Facility towards validation of thermal design of top shield for a pool type SFR has been conceived, constructed & commissioned. Detailed experiments were performed in this experimental facility having full-scale features. Steady state temperature distribution within the facility is measured for various heater plate temperatures in addition to simulating different operating states of the reactor. Following are the important observations (i) jet cooling system is effective in regulating the roof slab bottom plate temperature and thermal gradient across roof slab simulating normal operation of reactor, (ii) wire mesh insulation provided in roof slab-main vessel annulus is effective in obtaining gradual thermal gradient along main vessel top portion and inhibiting the setting up of cellular convection within annulus and (iii) cellular convection with four distinct convective cells sets in the annular gap between roof slab and small rotatable plug measuring ∼ϕ4 m in diameter & gap width varying from 16 mm to 30 mm. Repeatability of results is also ensured during all the above tests. The results presented in this paper is expected to provide reference data for validation of thermal hydraulic models in addition to serving as design validation of jet cooling system for pool type SFR.

Comparison of Pickering NGS performance with world power reactors, 1977

International Nuclear Information System (INIS)

Buhay, S.

Pickering NGS performance is compared, in highly graphic form, with the perfomance of other nuclear power plants around the world. The four Pickering reactors score in the top six, rated by gross capacity factor. Major system suppliers for world power reactors above 500 MW are cataloged. (E.C.B.)

Proceedings of the 2007 LWR Fuel Performance Meeting / TopFuel 2007 'Zero by 2010'

International Nuclear Information System (INIS)

2007-01-01

ANS, ENS, AESJ and KNS are jointly organizing the 2007 International LWR Fuel Performance Meeting following the successful ENS TopFuel meeting held during 22-26 October, 2006 in Salamaca, Spain. Merging three premier nuclear fuel design and performance meetings: the ANS LWR Fuel Performance Meeting, the ENS TopFuel and Asian Water Reactor Fuel Performance Meeting (WRFPM) created this international meeting. The meeting will be held annually on a tri-annual rotational basis in USA, Asia, and Europe. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as performance experience in commercial and test reactors. The meeting excludes front end and back end fuel issues, however, it covers all front and/or back issues that impact fuel designs and performance

University Reactor Instrumentation Grant. Final report 08/06/1998 - 08/13/1999

International Nuclear Information System (INIS)

Bajorek, S. M.

2000-01-01

A noble gas air monitoring system was purchased through the University Reactor Instrumentation Grant Program. This monitor was installed in the Kansas State TRIGA reactor bay at a location near the top surface of the reactor pool according to recommendation by the supplier. This system is now functional and has been incorporated into the facility license

Startup measurements on the CABRI reactor

International Nuclear Information System (INIS)

Kussmaul, G.; Bensoussan, P.; Dadillon, J.; Golinelli, C.; Tonolli, J.

1979-08-01

The CABRI reactor will be used for the investigation of the behavior of fresh and irradiated fast reactor fuel pins under TOP conditions. A startup programme has been carried out to measure fundamental data determining steady state and transient behavior of the driver core as well as data ensuring safe operation of the reactor. Special emphasis was laid on quantities not well known from previous neutronics calculations, e.g. prompt-neutron generation time, Doppler feedback and time-dependent reactivity injection. Utilizing the data inferred from measurements in the dynamic code DULCINEE good agreement between calculated and observed transient behavior of the driver core has been found

Tank type LMFBR type reactors

International Nuclear Information System (INIS)

Shimizu, Hiroshi

1985-01-01

Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

1982-01-01

A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

Effect of design parameters on performance of a top fired natural gas reformer

International Nuclear Information System (INIS)

Ebrahimi, Hadi; Mohammadzadeh, Jafar S. Soltan; Zamaniyan, Akbar; Shayegh, Flora

2008-01-01

A three-dimensional zone method was applied to an industrial fired heater of methane steam reforming reactor. Radiation heat transfer from all gases and surfaces inside the furnace was considered. Results from previous work and data of an industrial top fired furnace were used to validate the model. A maximum temperature in external reaction tube skin was obtained at about one third of the reactor length from top in the industrial furnace. Effect of important parameters such as emissivity, extinction coefficient, heat release pattern and flame angle on performance of the fired heater are presented. It was found that decreasing the extinction coefficients of combustion gases by 25% (from about 0.20 to 0.15) caused 2.6% rise in temperature of heat sink surfaces. It was demonstrated that the three-dimensional zone method developed in this work is simple, easy and flexible for modeling and simulation of the fired heaters

Fuel exchanger in FBR type reactor

International Nuclear Information System (INIS)

Shinden, Kazuhiko; Tanaka, Osamu.

1990-01-01

The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

Top reconstruction and boosted top experimental overview

CERN Document Server

Skinnari, Louise

2015-01-01

An overview of techniques used to reconstruct resolved and boosted top quarks is presented. Techniques for resolved top quark reconstruction include kinematic likelihood fitters and pseudo- top reconstruction. Many tools and methods are available for the reconstruction of boosted top quarks, such as jet grooming techniques, jet substructure variables, and dedicated top taggers. Different techniques as used by ATLAS and CMS analyses are described and the performance of different variables and top taggers are shown.

Assessment of proliferation resistances of aqueous reprocessing techniques using the TOPS methodology

International Nuclear Information System (INIS)

Åberg Lindell, M.; Grape, S.; Håkansson, A.; Jacobsson Svärd, S.

2013-01-01

Highlights: • Proliferation resistances of three possible LFR fuel cycles are assessed. • The TOPS methodology has been chosen for the PR assessment. • Reactor operation, reprocessing and fuel fabrication are examined. • Purex, Ganex, and a combination of Purex, Diamex and Sanex, are compared. • The safeguards analysis speaks in favor of Ganex as opposed to the Purex process. - Abstract: The aim of this study is to assess and compare the proliferation resistances (PR) of three possible Generation IV lead-cooled fast reactor fuel cycles, involving the reprocessing techniques Purex, Ganex and a combination of Purex, Diamex and Sanex, respectively. The examined fuel cycle stages are reactor operation, reprocessing and fuel fabrication. The TOPS methodology has been chosen for the PR assessment, and the only threat studied is the case where a technically advanced state diverts nuclear material covertly. According to the TOPS methodology, the facilities have been divided into segments, here roughly representing the different forms of nuclear material occurring in each examined fuel cycle stage. For each segment, various proliferation barriers have been assessed. The results make it possible to pinpoint where the facilities can be improved. The results show that the proliferation resistance of a fuel cycle involving recycling of minor actinides is higher than for the traditional Purex reprocessing cycle. Furthermore, for the purpose of nuclear safeguards, group actinide extraction should be preferred over reprocessing options where pure plutonium streams occur. This is due to the fact that a solution containing minor actinides is less attractive to a proliferator than a pure Pu solution. Thus, the safeguards analysis speaks in favor of Ganex as opposed to the Purex process

Gasification Performance of a Top-Lit Updraft Cook Stove

Directory of Open Access Journals (Sweden)

Yogesh Mehta

2017-10-01

Full Text Available This paper reports on an experimental study of a top-lit updraft cook stove with a focus on gasification. The reactor is operated with primary air only. The performance is studied for a variation in the primary airflow, as well as reactor geometry. Temperature in the reactor, air flow rate, fuel consumption rate, and producer gas composition were measured. From the measurements the superficial velocity, pyrolysis front velocity, peak bed temperature, air fuel ratio, heating value of the producer gas, and gasification rate were calculated. The results show that the producer gas energy content was maximized at a superficial velocity of 9 cm/s. The percent char remaining at the end of gasification decreased with increasing combustion chamber diameter. For a fixed superficial velocity, the gasification rate and producer gas energy content were found to scale linearly with diameter. The energy content of the producer gas was maximized at an air fuel (AF ratio of 1.8 regardless of the diameter.

FLUID MODERATED REACTOR

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1957-10-22

A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

2003-01-01

Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

Fuel motion in overpower tests of metallic integral fast reactor fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

1992-01-01

In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

A seismic performance and cost comparison of top and bottom supported liquid metal reactor vessels

International Nuclear Information System (INIS)

Carlson, T.M.; Kiciman, O.K.; Petrozelli, J.F.

1989-01-01

It is the premise of this paper that the revision of a pool LMR from a TSRV configuration to a specific bottom supported reactor vessel (BSRV) configuration can resolve the above TSRV disadvantages related to load path length and diversity, thereby improving seismic performance and simultaneously reducing RV block costs by reducing weights. This paper demonstrates this premise by comparing a reference TSRV block with a specific BSRV block design. Recent capital cost estimates ($/kWe) for U.S. liquid metal reactor (LMR) plant designs reveal that the balance of plant costs could be reduced below that of the balance of plant costs for a comparable light water reactor plant. However, in regions of high seismicity, non-seismically isolated LMR nuclear steam supply system weights are costs per kWe are two to three times the weights and costs of light water reactor nuclear steam supply systems. While all portions of the LMR nuclear steam supply system require examination for potential cost reductions, the focus of this paper is the reactor vessel (RV) block for a large pool plant

An investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a boiling water reactor anticipated transient without scram

International Nuclear Information System (INIS)

Peterson, C.E.; Chexal, V.K.; Gose, G.C.; Hentzen, R.D.; Layman, W.H.

1985-01-01

Under certain anticipated transient without scram (ATWS) sequences for a boiling water reactor, it would be desirable to reduce system power, particularly where the primary system has been isolated by closure of all main steam isolation valves and is discharging steam through its safety/relief valve system to the suppression pool. Reducing reactor power increases the time available to shut down the reactor by minimizing the heat dumped to the suppression pool and by helping to keep the suppression pool temperature within limits. Under proposed emergency procedure guidelines for the ATWS event, the reactor water level would be lowered to reduce reactor power. The analyses provide an assessment of the power level that would be attained, assuming the reactor operators were to reduce the the downcomer level down to the top of the active fuel

Development and investigation of the prestressed reinforced concrete vessels for the water cooled reactors in the FRG

International Nuclear Information System (INIS)

Medovikov, A.I.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Konevskij, V.N.

1980-01-01

An analysis of calculation results for characteristics of stress-strained state of reactor vessel made of prestressed reinforced concrete is presented. Experimental data obtained during the investigation into a model of reactor vessel top cover are given. Thermal shielding system both for boiling water and pressurized-water reactors has been considered and its working capacity has been evaluated. An analysis of experimental data show correctness of the method assumed for calculation of the reactor top cover which permits to exactly determine its stressed-strained state as well as the nature of crack propagation in the vessel and the structure supporting power. Ceramics is suggested to be used as a heat-insulating material

MEANS FOR SHIELDING AND COOLING REACTORS

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

Pressure suppression device for nuclear reactor building

International Nuclear Information System (INIS)

Ikegame, Noboru.

1992-01-01

In a nuclear reactor building, there are disposed cooling coils connected to an air supply duct at the outside of the building, an air supply blower, an air supply duct having the top end opened, an exhaustion duct having the top end opened and a bypassing pipeline interposed between the exhaustion duct and the air supply duct on the side of the inlet of the cooling coils. In the reactor building, when a radioactive material leakage accident should occur, an isolation valve is closed to isolate the building from the outside. Further, bypassing isolation valve is opened to form a closed cooling circuit by the cooling coils, the air supply blower and the air supply duct, the exhaustion duct and the bypassing pipeline in the reactor building. With such a constitution, since air as the atmosphere in the building is circulated through the closed cooling circuit and cooled by the cooling coils, the temperature is not elevated. Accordingly, since the pressure elevation of the atmosphere in the building is suppressed, the atmosphere containing radioactive materials do not flow out of the building. (I.N.)

In-vessel maintenance concepts for tokamak fusion reactors

International Nuclear Information System (INIS)

Kelly, V.P.; Berger, J.D.; Yount, J.A.

1983-01-01

Concepts for rail-mounted and guided in-vessel handling machines (IVM) for remote maintenance inside tokamak fusion reactors are described. The IVM designs are based on concepts for tethered remotely operated vehicles and feature the use of multiple manipulator arms for remote handling and remote-controlled TV cameras for remote viewing. The concepts include IVMs for both single or dual rail systems located in the top or bottom of the reactor vessel

Development of the cascade inertial-confinement-fusion reactor

International Nuclear Information System (INIS)

Pitts, J.H.

1985-01-01

Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li 2 O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO 2 granules. Average blanket exit temperature is 1670 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation

Development of the cascade inertial-confinement-fusion reactor

International Nuclear Information System (INIS)

Pitts, J.H.

1985-01-01

Cascade, originally conceived as a football-shaped, steel-walled reactor containing a Li 2 O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO 2 granules. Average blanket exit temperature is 1670 0 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation

Emergency reactor shutdown device

International Nuclear Information System (INIS)

Ikehara, Morihiko.

1982-01-01

Purpose: To smoothen the emergency operation of the control rod in a BWR type reactor and to eliminate the external discharge of radioactively contaminated water. Constitution: A drain receiving tank is connected through a scram valve to the top of a cylinder which is containing a hydraulic piston connected to a trombone-shaped control rod and an accumulator is connected through another scram valve to the bottom of the cylinder. The respective scram valves are constructed to be opened by the reactor emergency shutdown signal from a reactor control system in such a manner that drain valve and a vent valve of the tank normally opened at the standby time are closed after approx. 10 seconds from the opening of the scram valves. In this manner, back pressure is not applied to the hydraulic piston at the emergency time, thereby smoothly operating the control rod. (Sikiya, K.)

Safety aspects of LMR [liquid metal-cooled reactor] core design

International Nuclear Information System (INIS)

Cahalan, J.E.

1986-01-01

Features contributing to increased safety margins in liquid metal-cooled reactor (LMR) design are identified. The technical basis is presented for the performance of a pool-type reactor system with an advanced metallic alloy fuel in unprotected accidents. Results are presented from analyses of anticipated transients without scram, including loss-of-flow (LOF), transient overpower (TOP), and loss-of-heat-sink (LOHS) accidents

An automatic device for sample insertion and extraction to/from reactor irradiation facilities

International Nuclear Information System (INIS)

Alloni, L.; Venturelli, A.; Meloni, S.

1990-01-01

At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

Reactor head shielding apparatus

International Nuclear Information System (INIS)

Schukei, G.E.; Roebelen, G.J.

1992-01-01

This patent describes a nuclear reactor head shielding apparatus for mounting on spaced reactor head lifting members radially inwardly of the head bolts. It comprises a frame of sections for mounting on the lifting members and extending around the top central area of the head, mounting means for so mounting the frame sections, including downwardly projecting members on the frame sections and complementary upwardly open recessed members for fastening to the lifting members for receiving the downwardly projecting members when the frame sections are lowered thereto with lead shielding supported thereby on means for hanging lead shielding on the frame to minimize radiation exposure or personnel working with the head bolts or in the vicinity thereof

Ekspedeerimisfirmade TOP

Index Scriptorium Estoniae

2008-01-01

Ekspedeerimisfirmade TOP 57. Vt. samas: Tanel Raig. Majandus kukutab ekspedeerimisturgu. Diagramm: Väliskaubanduse statistika; Katrin Raie. Ekspedeerijad hakkavad rohkem koostööle rõhuma. Kommenteerib Jaan Lepp; Müügitulu TOP 10; Müügitulu kasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Eestis registreeritud Vene hiiglane; Ekspedeerimisturu kasumiliider kaotas 20 miljonit; Küsimustele vastab OÜ Contimer juht Dmitri Redkin

Reconstitutable fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Ferlan, S.J.; Kmonk, S.; Schallenberger, J.M.

1982-01-01

A reconstitutable fuel assembly for a nuclear reactor which includes a mechanical, rather than metallurgical, arrangement for connecting control rod guide thimbles to the top and bottom nozzles of a fuel assembly. Multiple sleeves enclosing control rod guide thimbles interconnect the top nozzle to the fuel assembly upper grid. Each sleeve is secured to the top nozzle by retaining rings disposed on opposite sides of the nozzle. Similar sleeves enclose the lower end of control rod guide thimbles and interconnect the bottom nozzle with the lowermost grid on the assembly. An end plug fitted in the bottom end of each sleeve extends through the bottom nozzle and is secured thereto by a retaining ring. Should it be necessary to remove a fuel rod from the assembly, the retaining rings in either the top or bottom nozzles may be removed to release the nozzle from the control rod guide thimbles and thus expose either the top or bottom ends of the fuel rods to fuel rod removing mechanisms

A top priority problem of national radiation protection - proper disposal of research reactor spent fuel

International Nuclear Information System (INIS)

Marinkovic, N.; Matausek, M.V.; Jovic, V.

1997-01-01

The paper presents basic facts about RA research reactor at the Vinca Institute. The present state of the RA reactor spent fuel storage pool appears to be a serious safety and radiological problem, which must be solved urgently, independent of the decision about the future status of the reactor itself. The following paragraphs describe current activities on improving storage conditions of the research reactor RA spent fuel. Activities performed so far, concerning identification and improvement of the spent fuel storage conditions are presented. These are verification of radiation protection measures, radiological and chemical analyses, visual inspection and photographing, safety analyses and nuclear criticality studies.A project for long-term solution of the research reactor spent fuel storage is proposed. In order to minimise further corrosion and establish strict control of all the relevant technological parameters of the utility, improvement of conditions for disposal of the fuel in the existing storage, is foreseen in the first phase. New dry storage for long-term storing of the spent fuel should be built during the second phase of the project. Particular attention is paid to the activities related to radiation protection and waste treatment, starting from standard monitoring and control, radiological analyses, regulations and legislation, to complicated handling of high level radioactive waste. (authors)

Experimental study of hot water layer in a model in scale of the Brazilian Multipurpose Reactor (RMB)

International Nuclear Information System (INIS)

Tomaz, Gabriel Caio Queiroz

2017-01-01

The Brazilian Multipurpose Reactor (RMB) is a 30 MW open pool research reactor planned to be constructed in Brazil. Such type of reactor is built inside a deep pool of purified and demineralized water, providing radiological protection still keeping the core accessible for maintenance and refueling. However, dissolved ions become activated in the pool water due to the core neutron flux, releasing radiation in the reactor room when the activated elements reach the top. Thus high power open pool reactors, as RMB, have an auxiliary thermal-hydraulic circuit that creates a Hot Water Layer (HWL) on the pool’s top, keeping the activated water under the HWL and mitigating the dose rate to which the operators are exposed to. The Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) built a 1/10 scale experimental bench of the RMB’s pool for the HWL investigation. This work presents the results of the pool’s heating due to the reactor startup in the HWL stability. (author)

Selective distribution of enzymes in a microfluidic reactor

DEFF Research Database (Denmark)

Daugaard, Anders Egede; Pereira Rosinha Grundtvig, Ines; Krühne, Ulrich

Off stoichiometric thiol-ene mixtures are well suited for preparation of microfluidic devices with highly functional surfaces. Here a two stage process employing first thiol-ene chemistry (TEC) to prepare two opposite parts of a microfluidic system with a 30x30 mm reactor and subsequently a thiol......-epoxy bonding was used to prepare a fully sealed microfluidic system. The reactor was surface functionalized in-situ with allyl glycidyl ether in different patterns (half-reactor, full-reactor, checkerboard structures) on the surface to provide a controlled distribution of epoxides. The method additionally...... enables the selective immobilization on either top-side or bottom-side or both sides of the reactor. Thereafter horseradish peroxidase was immobilized on the surface and activity tests illustrated how this distribution of the enzyme on the surface could be used to optimize the activity of the enzyme...

Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

International Nuclear Information System (INIS)

2010-01-01

The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

Top-down versus bottom-up processing of influence diagrams in probabilistic analysis

International Nuclear Information System (INIS)

Timmerman, R.D.; Burns, T.J.; Dodds, H.L. Jr.

1986-01-01

Recent work by Phillips and Selby has shown that influence diagram methodology can be a useful analytical tool in reactor safety studies. In some instances an influence diagram can be used as a graphical representation of probabilistic dependence within a system or event sequence. Under these circumstances, Bayesian statistics is employed to transform the relationships depicted in the influence diagram into the correct expression for a desired marginal probability (e.g. the top node). Top-down and bottom-up algorithms have emerged as the dominant methods for quantifying influence diagrams. The purpose of this paper is to demonstrate a potential error in employing the bottom-up algorithm when dealing with interdependencies

Rodded shutdown system for a nuclear reactor

International Nuclear Information System (INIS)

Golden, M.P.; Govi, A.R.

1978-01-01

A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature is described. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core

TopView - ATLAS top physics analysis package

CERN Document Server

Shibata, A

2007-01-01

TopView is a common analysis package which is widely used in the ATLAS top physics working group. The package is fully based on the official ATLAS software Athena and EventView and playing a central role in the collaborative analysis model. It is a functional package which accounts for a broad range issues in implementing physics analysis. As well as being a modular framework suitable as a common workplace for collaborators, TopView implements numerous analysis tools including a complete top-antitop reconstruction and single top reconstruction. The package is currently used to produce common ntuple from Monte Carlo production and future use cases are under rapid development. In this paper, the design and ideas behind TopView and the performance of the analyses implemented in the package are presented with detailed documentation of the contents and instruction for using the package.

Kinnisvarafirmade TOP 90

Index Scriptorium Estoniae

2002-01-01

TOP 90. Kinnisvara valdkondade TOP 5. Käibe TOP 30. Käibe kasvu TOP 30. Rentaabluse TOP 30. Kasumi TOP 30. Kasumi kasvu TOP 30. Varade tootlikkuse TOP 30. Kinnisvarafirmade üldandmed. Kinnisvarafirmade finantsandmed

Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

2003-01-01

The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

Simulated nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Berta, V.T.

1993-01-01

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end

Kinnisvarafirmade TOP 90

Index Scriptorium Estoniae

2003-01-01

Kinnisvarafirmade TOP 90. Rentaabluse TOP 30. Käibe TOP 30. Käibe kasvu TOP 30. Kasumi TOP 30. Kasumi kasvu TOP 30. Varade tootlikkuse TOP 30. Kinnisvarafirmade üldandmed. Kinnisvarafirmade finantsandmed

Physics of coolability of top flooded molten corium

International Nuclear Information System (INIS)

Kulkarni, P.P.; Singh, R.K.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

2011-01-01

During a postulated severe accident in a nuclear reactor in case of ex-vessel scenario the molten corium can be relocated in the containment cavity forming a melt pool. In order to arrest further progression of severe accident, complete quenching of the molten corium pool is necessary. Most common way to deal with ex-vessel scenario is to flood the melt pool with large quantity of water. However, the mechanism of coolability is much more complex involving multi-component, multiphase heat, mass and momentum transfer. In this paper, a mechanistic model has been presented for the corium coolability under top flooding conditions. The model has been validated with the experimental data of COMECO test facility available in literature. Simulations have been carried out using the model to explore the physics behind the corium coolability with MCCI under top flooding condition. Variations in the thermo-physical properties as a result of MCCI have been considered and its effect on coolability has been studied. (author)

Teedeehitusfirmade TOP 24

Index Scriptorium Estoniae

2006-01-01

Teedeehitusfirmade TOP. Vt. samas: Käibe TOP 10; Käibe kasvu TOP 10; Kasumi TOP 10; Kasumi Kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Teedeehitusfirmade üld- ja finantsandmed

Ehitusmaterjalitootjate TOP 95

Index Scriptorium Estoniae

2006-01-01

Ehitusmaterjalitootjate TOP. Vt. samas: Käibe TOP 10; Käibe kasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Ehitusmaterjalitootjate üld- ja finantsandmed

Koolitusfirmade TOP 50

Index Scriptorium Estoniae

2006-01-01

Koolitusfirmade TOP. Vt. samas: Käibe TOP 10; Käibe kasvu TOP 10; Majandustegevuse kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Koolitusfirmade üld- ja finantsandmed

Põllumajandustootjate TOP 50

Index Scriptorium Estoniae

2006-01-01

Põllumajandustootjate TOP. Vt. samas: Käibe TOP 10; Käibe kasvu TOP 10; Majandustegevuse kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Põllumajandustootjate üld- ja finantsandmed

Numerical analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

International Nuclear Information System (INIS)

Schweizer, Fernando Lage Araújo

2014-01-01

The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)

Ageing study of Cirus reactor vessel expansion bellow

International Nuclear Information System (INIS)

Ramana, W.V.; Dutta, B.K.; Kushwaha, H.S.; Sahu, A.K.; Bhatnagar, A.; Pant, R.C.

1994-01-01

Expansion bellow of Cirus reactor vessel is a comparatively weak component which is joined to top tube sheet and shell by helium tight lap weld. This has been subjected to thermal stress caused by high temperature during reactor operation and thermal shock due to trip or shutdown. Therefore a finite element analysis was carried out to assess thermal stresses and fatigue life of the component. It was found that the fluctuating stress in the bellow is far less than its endurance limit. (author). 2 tabs., 3 figs

A transparent Pyrex μ-reactor for combined in situ optical characterization and photocatalytic reactivity measurements

International Nuclear Information System (INIS)

Dionigi, F.; Hansen, O.; Nielsen, M. G.; Chorkendorff, I.; Vesborg, P. C. K.; Pedersen, T.

2013-01-01

A new Pyrex-based μ-reactor for photocatalytic and optical characterization experiments is presented. The reactor chamber and gas channels are microfabricated in a thin poly-silicon coated Pyrex chip that is sealed with a Pyrex lid by anodic bonding. The device is transparent to light in the UV-vis-near infrared range of wavelengths (photon energies between ∼0.4 and ∼4.1 eV). The absorbance of a photocatalytic film obtained with a light transmission measurement during a photocatalytic reaction is presented as a proof of concept of a photocatalytic reactivity measurement combined with in situ optical characterization. Diffuse reflectance measurements of highly scattering photocatalytic nanopowders in a sealed Pyrex μ-reactor are also possible using an integrating sphere as shown in this work. These experiments prove that a photocatalyst can be characterized with optical techniques after a photocatalytic reaction without removing the material from the reactor. The catalyst deposited in the cylindrical reactor chamber can be illuminated from both top and bottom sides and an example of application of top and bottom illumination is presented

Event group importance measures for top event frequency analyses

International Nuclear Information System (INIS)

1995-01-01

Three traditional importance measures, risk reduction, partial derivative, nd variance reduction, have been extended to permit analyses of the relative importance of groups of underlying failure rates to the frequencies of resulting top events. The partial derivative importance measure was extended by assessing the contribution of a group of events to the gradient of the top event frequency. Given the moments of the distributions that characterize the uncertainties in the underlying failure rates, the expectation values of the top event frequency, its variance, and all of the new group importance measures can be quantified exactly for two familiar cases: (1) when all underlying failure rates are presumed independent, and (2) when pairs of failure rates based on common data are treated as being equal (totally correlated). In these cases, the new importance measures, which can also be applied to assess the importance of individual events, obviate the need for Monte Carlo sampling. The event group importance measures are illustrated using a small example problem and demonstrated by applications made as part of a major reactor facility risk assessment. These illustrations and applications indicate both the utility and the versatility of the event group importance measures

Event group importance measures for top event frequency analyses

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-31

Three traditional importance measures, risk reduction, partial derivative, nd variance reduction, have been extended to permit analyses of the relative importance of groups of underlying failure rates to the frequencies of resulting top events. The partial derivative importance measure was extended by assessing the contribution of a group of events to the gradient of the top event frequency. Given the moments of the distributions that characterize the uncertainties in the underlying failure rates, the expectation values of the top event frequency, its variance, and all of the new group importance measures can be quantified exactly for two familiar cases: (1) when all underlying failure rates are presumed independent, and (2) when pairs of failure rates based on common data are treated as being equal (totally correlated). In these cases, the new importance measures, which can also be applied to assess the importance of individual events, obviate the need for Monte Carlo sampling. The event group importance measures are illustrated using a small example problem and demonstrated by applications made as part of a major reactor facility risk assessment. These illustrations and applications indicate both the utility and the versatility of the event group importance measures.

Radionuclide trap for liquid metal cooled reactors

International Nuclear Information System (INIS)

McGuire, J.C.; Brehm, W.F.

1978-10-01

At liquid metal cooled reactor operating temperatures, radioactive corrosion product transport and deposition in the primary system will be sufficiently high to limit access time for maintenance of system components. A radionuclide trap has been developed to aid in controlling radioactivity transport. This is a device which is located above the reactor core and which acts as a getter, physically immobilizing radioactive corrosion products, particularly 54 Mn. Nickel is the getter material used. It is most effective at temperatures above 450 0 C and effectiveness increases with increasing temperature. Prototype traps have been tested in sodium loops for 40,000 hours at reactor primary temperatures and sodium velocities. Several possible in-reactor trap sites were considered but a location within the top of each driver assembly was chosen as the most convenient and effective. In this position the trap is changed each time fuel is changed

Autotranspordifirmade TOP 100

Index Scriptorium Estoniae

2006-01-01

Ilmunud ka: Delovõje Vedomosti : Transport i Logistika 29. nov. lk. 10-11. Autofirmade TOP 100. Vt. samas: Käibe TOP 10; Käibe kasvu TOP 10; Majandustegevuse kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Autotranspordifirmade üld- ja finantsandmed. Delovõje Vedomosti : Transport i Logistika sisaldab tabelit Autofirmade TOP 100

Top-down versus bottom-up processing of influence diagrams in probabilistic analysis

International Nuclear Information System (INIS)

Timmerman, R.D.; Burns, T.J.; Dodds, H.L. Jr.

1984-01-01

Recent work by Phillips et al., and Selby et al., has shown that influence diagram methodology can be a useful analytical tool in reactor safety studies. An influence diagram is a graphical representation of probabilistic dependence within a system or event sequence. Bayesian statistics are employed to transform the relationships depicted in the influence diagram into the correct expression for a desired marginal probability (e.g. the top event). As with fault trees, top-down and bottom-up algorithms have emerged as the dominant methods for quantifying influence diagrams. Purpose of this paper is to demonstrate a potential error in employing the bottom-up algorithm when dealing with interdependencies. In addition, the computing efficiency of both methods is discussed

Study of top and anti-top mass difference

CERN Document Server

Leedumrongwatthanakun, Saroch

2013-01-01

The invariance of the standard model under CPT transformations leads to the equality of particle and antiparticle masses. The recent measurements performed by the CMS experiment on the top anti-top mass difference are a test of such symmetry. In this work non-perturbative QCD effects, which may eventually lead to an apparent difference in the mass of a top and anti-top quark, are studied.

Core access system for nuclear reactor

International Nuclear Information System (INIS)

Andrea, C.

1977-01-01

Disclosed is an improved nuclear reactor arrangement to facilitate both through-the-head instrumentation and insertion and removal of assemblies from the nuclear core. The arrangement is of the type including a reactor vessel head comprising a large rotatable cover having a plurality of circular openings therethrough, a plurality of upwardly extending nozzles mounted on the upper surface of a large cover, and a plurality of upwardly extending skirts mounted on a large cover about the periphery or boundary of the circular openings; a plurality of small plugs for each of the openings in the large cover, the plugs also having nozzles mounted on the upper surface thereof, and drive mechanisms mounted on top of some of the nozzles and having means extending therethrough into the reactor vessel, the drive mechanisms and nozzles extending above the elevation of the upwardly extending skirts

Particle bed reactor nuclear rocket concept

International Nuclear Information System (INIS)

Ludewig, H.

1991-01-01

The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

Energy Technology Data Exchange (ETDEWEB)

GERBER MS

2009-04-28

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

Cooling nuclear reactor fuel

International Nuclear Information System (INIS)

Porter, W.H.L.

1975-01-01

Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

Reliability and results, but little room to manoeuvre

International Nuclear Information System (INIS)

Atkinson, Ian

1994-01-01

A wide variety of research programs to produce cost-effective and reliable inspection techniques have arisen following the discovery of stress corrosion cracks in the control rod drive mechanism of pressurized water reactors, notably in France, Belgium and Spain. This article describes the research program results from a cooperative partnership between Comex Nucleaire, Westinghouse Electric and AEA Technology. The package developed offers techniques to provide complete capability in virtually all the design configurations used world-wide. After extensive acceptance trials in France and the United States the techniques are now being used on site at Bugey 3. (UK)

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

Alternative water injection device to reactor equipment facility

International Nuclear Information System (INIS)

Yamashita, Masahiro.

1995-01-01

The device of the present invention injects water to the reactor and the reactor container continuously for a long period of time for preventing occurrence of a severe accident in a BWR type reactor and maintaining the integrity of the reactor container even if the accident should occur. Namely, diesel-driven pumps disposed near heat exchangers of a reactor after-heat removing system (RHR) are operated before the reactor is damaged by the after heat to cause reactor melting. A sucking valve disposed to a pump sucking pipeline connecting a secondary pipeline of the RHR heat exchanger and the diesel driving pump is opened. A discharge valve disposed to a pump discharge pipeline connecting a primary pipeline of the RHR heat exchanger and the diesel driving pump is opened. With such procedures, sea water is introduced from a sea water taking port through the top end of the secondary pipeline of the RHR heat exchanger and water is injected into the inside of the pressure vessel or the reactor container by way of the primary pipeline of the RHR heat exchanger. As a result, the reactor core is prevented from melting even upon occurrence of a severe accident. (I.S.)

50 pc of nuclear: 5 nuclear plants to shut down as a priority

International Nuclear Information System (INIS)

2013-01-01

Whereas only the shutting down of Fessenheim has been evoked until now, and as the objective of reduction of production of nuclear electricity implies the shutting down of 20 reactors by 2020, this document reports the study of five nuclear plants to be possibly shut down (Blayais, Bugey, Fessenheim, Gravelines, and Tricastin). For each plant, the safety level has been assessed (with respect to installation age, size, reactor power, fuel type, condition of primary circuits, basement type, situation of cooling pools), the risks of natural and non natural external aggressions have been assessed (flooding, seismic risk, air aggression, industrial risk, fire risk, landslide risk), and the consequences of an accident have been assessed regarding contamination diffusion, border or town vicinity, population density and evacuation requirements and constraints, and social and economic configuration

A resting bottom sodium cooled fast reactor

International Nuclear Information System (INIS)

Costes, D.

2012-01-01

This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

Measuring top-quark polarization in top-pair + missing-energy events.

Science.gov (United States)

Berger, Edmond L; Cao, Qing-Hong; Yu, Jiang-Hao; Zhang, Hao

2012-10-12

The polarization of a top quark can be sensitive to new physics beyond the standard model. Since the charged lepton from top-quark decay is maximally correlated with the top-quark spin, it is common to measure the polarization from the distribution in the angle between the charged lepton and the top-quark directions. We propose a novel method based on the charged lepton energy fraction and illustrate the method with a detailed simulation of top-quark pairs produced in supersymmetric top squark pair production. We show that the lepton energy ratio distribution that we define is very sensitive to the top-quark polarization but insensitive to the precise measurement of the top-quark energy.

Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

Energy Technology Data Exchange (ETDEWEB)

McKenzie, J L [Armed Forces Radiobiology Research Institute, Bethesda, MD (United States)

1974-07-01

The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the {sup 41}Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

International Nuclear Information System (INIS)

McKenzie, J.L.

1974-01-01

The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the 41 Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

Device for extracting steam or gas from the primary coolant line leading from a reactor pressure vessel to a straight through boiler or from the top primary boiler chamber of a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Schatz, K.

1982-01-01

In such a nuclear reactor, a steam or gas cushion can form when the primary system is refilled, which can cause blocking of the natural circulation or filling of the system in the area of the hot primary coolant pipe or in the top primary boiler chamber. In order to remove such a steam or gas cushion, a ventilation pipe starting from the bend of the primary coolant line is connected to the feed pipe for introducing water into the primary system. The feed pipe is designed on the principle of the vacuum pump in the area of the opening of the ventilation pipe. There is a sub-pressure in the ventilation pipe, which makes it possible to extract the steam or gas. After mixing in the area of the opening, the steam condenses or is distributed with the gas in the primary coolant. (orig.) [de

VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

International Nuclear Information System (INIS)

Doshi, P.K.; DeMario, E.E.; Knott, R.P.

1993-01-01

Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

Top quark pair production and top quark properties at CDF

Energy Technology Data Exchange (ETDEWEB)

Moon, Chang-Seong [INFN, Pisa

2016-06-02

We present the most recent measurements of top quark pairs production and top quark properties in proton-antiproton collisions with center-of-mass energy of 1.96 TeV using CDF II detector at the Tevatron. The combination of top pair production cross section measurements and the direct measurement of top quark width are reported. The test of Standard Model predictions for top quark decaying into $b$-quarks, performed by measuring the ratio $R$ between the top quark branching fraction to $b$-quark and the branching fraction to any type of down quark is shown. The extraction of the CKM matrix element $|V_{tb}|$ from the ratio $R$ is discussed. We also present the latest measurements on the forward-backward asymmetry ($A_{FB}$) in top anti-top quark production. With the full CDF Run II data set, the measurements are performed in top anti-top decaying to final states that contain one or two charged leptons (electrons or muons). In addition, we combine the results of the leptonic forward-backward asymmetry in $t\\bar t$ system between the two final states. All the results show deviations from the next-to-leading order (NLO) standard model (SM) calculation.

Study on thermodynamic cycle of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Qu Xinhe; Yang Xiaoyong; Wang Jie

2017-01-01

The development trend of the (very) High temperature gas-cooled reactor is to gradually increase the reactor outlet temperature. The different power conversion units are required at the different reactor outlet temperature. In this paper, for the helium turbine direct cycle and the combined cycle of the power conversion unit of the High temperature gas-cooled reactor, the mathematic models are established, and three cycle plans are designed. The helium turbine direct cycle is a Brayton cycle with recuperator, precooler and intercooler. In the combined cycle plan 1, the topping cycle is a simple Brayton cycle without recuperator, precooler and intercooler, and the bottoming cycle is based on the steam parameters (540deg, 6 MPa) recommended by Siemens. In the combined cycle plan 2, the topping cycle also is a simple Brayton cycle, and the bottoming cycle which is a Rankine cycle with reheating cycle is based on the steam parameters of conventional subcritical thermal power generation (540degC, 18 MPa). The optimization results showed that the cycle efficiency of the combined cycle plan 2 is the highest, the second is the helium turbine direct cycle, and the combined cycle plan 2 is the lowest. When the reactor outlet temperature is 900degC and the pressure ratio is 2.02, the cycle efficiency of the combined cycle plan 2 can reach 49.7%. The helium turbine direct cycle has a reactor inlet temperature above 500degC due to the regenerating cycle, so it requires a cooling circuit for the internal wall of the reactor pressure vessel. When the reactor outlet temperature increases, the increase of the pressure ratio required by the helium turbine direct cycle increases may bring some difficulties to the design and manufacture of the magnetic bearings. For the combined cycle, the reactor inlet temperature can be controlled below than 370degC, so the reactor pressure vessel can use SA533 steel without cooling the internal wall of the reactor pressure vessel. The pressure

MCNP evaluation of top node control rod depletion below the core in KKL

International Nuclear Information System (INIS)

Beran, Tâm; Seltborg, Per; Lindahl, Sten-Örjan; Bieli, Roger; Ledergerber, Guido

2014-01-01

In previous studies, there has been identified a significant discrepancy in the BWR control rod top node depletion between the two core simulator nodal codes POLCA7 and PRESTO-2, which indicates that there is a large general uncertainty in nodal codes in calculating the top node depletion of fully withdrawn control rods. In this study, the stochastic Monte Carlo code MCNP has been used to calculate the top node control rod depletion for benchmarking the nodal codes. By using the TIP signal obtained from an extended TIP campaign below the core performed in the KKL reactor, the MCNP model has been verified by comparing the axial profile between the TIP data and the gamma flux calculated by MCNP. The MCNP results have also been compared with calculations from POLCA7, which was found to yield slightly higher depletion rates than MCNP. It was also found that the 10 B depletion in the top node is very sensitive to the exact axial location of the control rod top when it is fully withdrawn. By using the MCNP results, the neutron flux model below the core in the nodal codes can be improved by implementing an exponential function for the neutron flux. (author)

Stresses in transition region of VVER-1000 reactor vessels

International Nuclear Information System (INIS)

Namgung, I.; Nguye, T.L.

2014-01-01

Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

Stresses in transition region of VVER-1000 reactor vessels

Energy Technology Data Exchange (ETDEWEB)

Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Nguye, T.L. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); National Research Inst. of Mechanical Engineering, Hanoi City, Vietnam (China)

2014-07-01

Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

Stack Monitoring System At PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Zamrul Faizad Omar; Mohd Sabri Minhat; Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Izhar Abu Hussin

2014-01-01

This paper describes the current Stack Monitoring System at PUSPATI TRIGA Reactor (RTP) building. A stack monitoring system is a continuous air monitor placed at the reactor top for monitoring the presence of radioactive gaseous in the effluent air from the RTP building. The system consists of four detectors that provide the reading for background, particulate, Iodine and Noble gas. There is a plan to replace the current system due to frequent fault of the system, thus thorough understanding of the current system is required. Overview of the whole system will be explained in this paper. Some current results would be displayed and moving forward brief plan would be mentioned. (author)

Microbial community composition of a down-flow hanging sponge (DHS) reactor combined with an up-flow anaerobic sludge blanket (UASB) reactor for the treatment of municipal sewage.

Science.gov (United States)

Kubota, Kengo; Hayashi, Mikio; Matsunaga, Kengo; Iguchi, Akinori; Ohashi, Akiyoshi; Li, Yu-You; Yamaguchi, Takashi; Harada, Hideki

2014-01-01

The microbial community composition of a down-flow hanging sponge (DHS) reactor in an up-flow anaerobic sludge blanket (UASB)-DHS system used for the treatment of municipal sewage was investigated. The clone libraries showed marked differences in microbial community composition at different reactor heights and in different seasons. The dominant phylotypes residing in the upper part of the reactor were likely responsible for removing organic matters because a significant reduction in organic matter in the upper part was observed. Quantification of the amoA genes revealed that the proportions of ammonia oxidizing bacteria (AOB) varied along the vertical length of the reactor, with more AOB colonizing the middle and lower parts of the reactor than the top of the reactor. The findings indicated that sewage treatment was achieved by a separation of microbial habitats responsible for organic matter removal and nitrification in the DHS reactor. Copyright © 2013 Elsevier Ltd. All rights reserved.

Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

Energy Technology Data Exchange (ETDEWEB)

Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

2012-11-15

One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

. Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

International Nuclear Information System (INIS)

Yusuf, I; Mati, A. A.

2010-01-01

The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.

Design of the Demineralized Water Make-up Line to Maintain the Normal Pool Water Level of the Reactor Pool in the Research Reactor

International Nuclear Information System (INIS)

Yoon, Hyun Gi; Choi, Jung Woon; Yoon, Ju Hyeon; Chi, Dae Young

2012-01-01

In many research reactors, hot water layer system (HWLS) is used to minimize the pool top radiation level. Reactor pool divided into the hot water layer at the upper part of pool and the cold part below the hot water layer with lower temperature during normal operation. Water mixing between these layers is minimized because the hot water layer is formed above cold water. Therefore the hot water layer suppresses floatation of cold water and reduces the pool top radiation level. Pool water is evaporated form the surface to the building hall because of high temperature of the hot water layer; consequently the pool level is continuously fallen. Therefore, make-up water is necessary to maintain the normal pool level. There are two way to supply demineralized water to the pool, continuous and intermittent methods. In this system design, the continuous water make-up method is adopted to minimize the disturbance of the reactor pool flow. Also, demineralized water make-up is connected to the suction line of the hot water layer system to raise the temperature of make-up water. In conclusion, make-up demineralized water with high temperature is continuously supplied to the hot water layer in the pool

Telekommunikatsiooni TOP aastal 2003

Index Scriptorium Estoniae

2004-01-01

Telekommunikatsiooni TOP aastal 2003. Käibe TOP 10. Käibe kasvu TOP 10. Rentaabluse TOP 10. Kasumi TOP 10. Kasumi kasvu TOP 10. Omakapitali tootlikkuse TOP 10. Telekommunikatsioonifirmade üldandmed. Telekommunikatsioonifirmade finantsandmed

Application of Candle burnup to small fast reactor

International Nuclear Information System (INIS)

Sekimoto, H.; Satoshi, T.

2004-01-01

A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. An equilibrium state was obtained for a large fast reactor (core radius is 2 m and reflector thickness is 0.5 m) successfully by using a newly developed direct analysis code. However, it is difficult to apply this burnup strategy to small reactors, since its neutron leakage becomes large and neutron economy becomes worse. Fuel enrichment should be increased in order to sustain the criticality. However, higher enrichment of fresh fuel makes the CANDLE burnup difficult. We try to find some small reactor designs, which can realize the CANDLE burnup. We have successfully find a design, which is not the CANDLE burnup in the strict meaning, but satisfies qualitatively its characteristics mentioned at the top of this abstract. In the final paper, the general description of CANDLE burnup and some results on the obtained small fast reactor design are presented.(author)

Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

Science.gov (United States)

Peterson, Per F.

2013-05-14

A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

Jaekaubandusettevõtete TOP 70

Index Scriptorium Estoniae

2005-01-01

Jaekaubandusettevõtete TOP 70; Käibe TOP 25; Kasumi TOP 25; Käibe kasvu TOP 20; Kasumi kasvu TOP 20; Rentaabluse TOP 20; Omakapitali tootlikkuse TOP 20; Jaekaubandusettevõtete üld- ja finantsandmed

Pump/heat exchanger assembly for pool-type reactor

International Nuclear Information System (INIS)

Nathenson, R.D.; Slepian, R.M.

1987-01-01

A heat exchanger and pump assembly comprising a heat exchanger including a housing for defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. A pump is disposed beneath the heat exchanger and is comprised of a plurality of flow couplers disposed in a circular array. Each flow coupler is comprised of a pump duct for receiving a first electrically conductive fluid, i.e. the primary liquid metal, from a pool thereof, and a generator duct for receiving a second electrically conductive fluid, i.e. the intermediate liquid metal. The primary liquid metal is introduced from the reactor pool into the top, inlet ends of the tubes, flowing downward therethrough to be discharged from the tubes' bottom ends directly into the reactor pool. The primary liquid metal is variously introduced into the pump ducts directly from the reactor pool, either from the bottom or top end of the flow coupler. The intermediate fluid introduced into the generator ducts via the inlet duct and inlet plenum and after leaving the generator ducts passes through the annular cavity of the exchanger to cool the primary liquid in the tubes. The annular magnetic field of the pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of the intermediate metal. (author)

Biological mine water treatment operating a one stage reactor system

CSIR Research Space (South Africa)

Baloyi, MJ

2006-05-01

Full Text Available rumen fluid as source of the fermentation organisms, were utilised as electron donor when sulphate, as the electron acceptor, is converted to sulphide. The feed water entered the reactor at the top, to allow the water to get in contact with grass...

Safety requirements in the design of research reactors: A Canadian perspective

International Nuclear Information System (INIS)

Lee, A.G.; Langman, V.J.

2000-01-01

In Canada, the formal development of safety requirements for the design of research reactors in general began under an inter-organizational Small Reactor Criteria Committee. This committee developed safety and licensing criteria for use by several small reactor projects in their licensing discussions with the Atomic Energy Control Board. The small reactor projects or facilities represented included the MAPLE-X10 reactor, the proposed SES-10 heating reactor and its prototype, the SDR reactor at the Whiteshell Laboratories, the Korea Multipurpose Research Reactor (a.k.a., HANARO) in Korea, the SCORE project, and the McMaster University Nuclear Reactor. The top level set of criteria which form a safety philosophy and serve as a framework for more detailed developments was presented at an IAEA Conference in 1989. AECL continued this work to develop safety principles and design criteria for new small reactors. The first major application of this work has been to the design, safety analysis and licensing of the MAPLE 1 and 2 reactors for the MDS Nordion Medical Isotope Reactor Project. This paper provides an overview of the safety principles and design criteria. Examples of an implementation of these safety principles and design criteria are drawn from the work to design the MAPLE 1 and 2 reactors. (author)

BWR Mark I pressure suppression study: characterization of the vertical load function utilizing bench top model tests

International Nuclear Information System (INIS)

McCauley, E.W.; Lai, W.

1977-02-01

A study was conducted to characterize the mechanisms which give rise to observed oscillations in the vertical load function (VLF) of bench top pool dynamics tests. This is part of a continuing investigation at the Lawrence Livermore Laboratory of the General Electric Mark I Nuclear Reactor pressure suppression system

Failure analysis of carbide fuels under transient overpower (TOP) conditions

International Nuclear Information System (INIS)

Nguyen, D.H.

1980-06-01

The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin

Ekspedeerimisettevõtete TOP 50

Index Scriptorium Estoniae

2006-01-01

Ilmunud ka: Delovõje Vedomosti : Transport i Logistika nr. 11, 29. nov. lk. 32-35. Ekspedeerimisettevõtete TOP. Vt. samas: Käibe TOP 10; Käibekasvu TOP 10; Kasumi TOP 10; Kasumikasvu TOP 10; Rentaabluse TOP 10; ROE TOP 10; Ekspedeerimisettevõtete üld- ja finantsandmed. Ajal. Delovõje Vedomosti : Transport i Logistika toodud ainult Ekspedeerimisettevõtete TOP 50

Audiitorfirmade TOP aastal 2007

Index Scriptorium Estoniae

2008-01-01

Audiitorfirmade TOP 51. Vt. samas: Urve Vilk. Audiitoriteni pole majanduslangus jõudnud; Intervjuu I.S. Audiitorteenuste OÜ omaniku Irina Somovaga; Käibe TOP 10; Käibe kasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10

Koolitusfirmade TOP aastal 2006

Index Scriptorium Estoniae

2007-01-01

Koolitusfirmade TOP. Vt. samas: Käibe TOP10; Käibe kasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Varade tootlikkuse TOP 10; Pille Rõivas. Võtmetegur meeskond; Vain & Partnerid: uudsus peitub sisus; Kristo Kiviorg. Teel Baltimaade koolitajate tippu

Recent advances in reactor protection and control system technology

International Nuclear Information System (INIS)

Anon.

1997-01-01

After a first-generation digital integrated protection system has been installed on all 1300 MWe PWR units in France, a new digital protection system was developed for the 1450 MWe units, using local area networks, fiber optics, Motorola 68000 microprocessors, and a modular design allowing for the design of any system on the basis of around 50 types of standard cards. In 1993, an upgrading program for this equipment was launched in order to reduce costs, in particular software development costs, further improve hardware modularity and facilitate integration and connection to existing equipment. The basic principles of the units are described together with the implementation of computer-aided software engineering (CASE) tools, interfaces with hard-wired equipment, and multiplexed connections. The nuclear instrumentation systems at the Fessenheim and Bugey plants have been renovated with these equipment

Autotranspordi TOP aastal 2007

Index Scriptorium Estoniae

2008-01-01

TOP 50. Vt. samas: Käibe TOP 10; Käibe kasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Marika Roomere. Täisteenuse pakkumine kergitas tulemusi; Jupiter Plus otsib järjest uusi kasvuvõimalusi; EST-Trans Kaubaveod teenib kasumit toiduvedamisega

Measurement of the mixing leptonic parameter θ13 at the Double Chooz reactor antineutrino experiment

International Nuclear Information System (INIS)

Durand, V.

2012-01-01

The Double Chooz experiment aims at measuring the neutrino mixing parameter θ13 by studying the oscillations of de ν-bar e produced by the Chooz nuclear reactors located in France. The experimental concept consists in comparing the signal of two identical 10.3 m 3 detectors, allowing to cancel most of the experimental systematic uncertainties. The near detector, whose goal is the flux normalization and a measurement without oscillation, is expected to be delivered in 2013. The farthest detector from the source is taking data since April 2011 and is sensitive to θ 13 , which is expected to affect both the rate and the shape of the measured de ν-bar e . In this thesis, are first presented the Double Chooz experiment, with its ν-bar e source, its detection method, and the expected signal and backgrounds. In order to perform a selection, important quantities have to be reconstructed, calibrated, and saved in data files. The channel time offsets determination, the energy and vertex reconstruction algorithm CocoReco, the reconstruction packages of the Common Trunk, and the light trees maker Cheetah are especially presented. Concerning the data analysis, all the selection cuts and results for signal and backgrounds are discussed, particularly the multiplicity cut, the multiple off time window method, the lithium veto cut, and the cosmogenic 9 Li background studies. The Double Chooz experiment observed 8,249 de ν-bar e candidates in 227.93 days in its far detector only. The reactor antineutrino flux prediction used the Bugey 4 flux measurement after correction for differences in core composition. The expectation in case of no-oscillation is 8,937 events and this deficit is interpreted as evidence for ν-bar e disappearance. From a rate and shape analysis, is found sin 2 2θ = 0,109± 0,030 (stat) ± 0,025 (syst), with Δm 2 31 = 2,32 x 10 -3 eV 2 , while the no-oscillation hypothesis is even excluded at 2.9 σ. (author) [fr

Koolitusfirmade TOP aastal 2007

Index Scriptorium Estoniae

2008-01-01

Koolitusfirmade TOP. Vt. samas: Käibe TOP 10; Käibekasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Signe Sillasoo. Invicta tahab lähiaastail laieneda Eestis ja mujalgi; Ketlin Priilinn. Addenda kasutas ära majanduse soodsa seisu. Kommenteerib Heli Sõmer. Juhtide hoiakute muutmisega tõus esikolmikusse

Top quark property measurements in single top

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00386283; The ATLAS collaboration

2016-01-01

A review of the recent results on measurements of top quark properties in single top quark processes, performed at the LHC by ATLAS and CMS is presented. The measurements are in good agreement with predictions and no deviations from Standard Model expectations have been observed.

Design requirements of instrumentation and control systems for next generation reactor

International Nuclear Information System (INIS)

Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju

1994-03-01

In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator's aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs

Design requirements of instrumentation and control systems for next generation reactor

Energy Technology Data Exchange (ETDEWEB)

Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-03-01

In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator`s aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs.

Results on top-quark physics and top-quark-like signatures by CMS

Science.gov (United States)

Chabert, Eric; CMS Collaboration

2017-07-01

This report reviews the results obtained by the CMS Collaboration on top quark physics, focusing on the latest ones based on p-p collisions provided by the LHC at \\sqrt{s}=13{{TeV}} during Run II. It covers measurements of single-top, top quark pairs and associated productions as well as measurements of top quark properties. Finally several beyond the standard model searches involving top quark in the final states are presented, such as searches for supersymmetry in the third generation, heavy resonances decaying into a top quark pair, or dark matter produced in association to a single-top or a top quark pair.

Emergency core cooling system for a fast reactor

International Nuclear Information System (INIS)

Johnson, H.G.; Madsen, R.N.

1976-01-01

The main heat transport system for a liquid-metal-cooled nuclear reactor is constructed with elevated piping and guard vessels or pipes around all components of the system below the elevation of the elevated piping so the head developed by the pumps at emergency motor speed will be unsufficient to lift the liquid-metal-coolant over the top of the guard tanks or pipes or out of the elevated piping in the event of a loss-of-coolant accident. In addition, inlet downcomers to the reactor vessel are contained within guard standpipes having a clearance volume as small as practicable. 4 claims, 2 drawing figures

Approach to developing reliable space reactor power systems

International Nuclear Information System (INIS)

Mondt, J.F.; Shinbrot, C.H.

1991-01-01

The Space Reactor Power System Project is in the engineering development phase of a three-phase program. During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described in this paper along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top down systems approach which includes a point design based on a detailed technical specification of a 100 kW power system

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

International Nuclear Information System (INIS)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

Radiation shield for nuclear reactors

International Nuclear Information System (INIS)

Weissenfluh, J.A.

1980-01-01

A reusable radiation shield for use in a reactor installation comprises a thin-walled, flexible and resilient container, made of plastic or elastomeric material, containing a hydrogenous fluid with boron compounds in solution. The container can be filled and drained in position and the fluid can be recirculated if required. When not in use the container can be folded and stored in a small space. The invention relates to a shield to span the top of the annular space between a reactor vessel and the primary shield. For this purpose a continuous toroidal container or a series of discrete segments is used. Other forms can be employed for different purposes, e.g. mattress- or blanket-like forms can be draped over potential sources of radiation or suspended from a mobile carrier and placed between a worker and a radiation source. (author)

Status of the top quark: Top production cross section and top properties

Energy Technology Data Exchange (ETDEWEB)

Boisvert, V.; /Rochester U.

2006-08-01

This report describes the latest cross section and property measurements associated with the top quark at the Tevatron Run II. The largest data sample used is 760 pb{sup -1} of integrated luminosity. Due to its large mass, the top quark might be involved in the process of electroweak symmetry breaking, making it a useful probe for signs of new physics.

Reisifirmade TOP 40 aastal 2002

Index Scriptorium Estoniae

2003-01-01

Reisifirmade TOP 40 aastal 2002. Reisifirmade TOP-i esikümme. Käibe TOP 40. Kasumi TOP 40. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Reisifirmade üldandmed. Reisifirmade finantsandmed. Tehnilise käibe alusel arvutatud edetabelid: Reisifirmade TOP 25; Käibe TOP 40; Rentaabluse TOP 10; Käibe kasvu TOP 10

The dynamic pressure measurements of the nuclear reactor coolant for condition-based maintenance of the reactor

International Nuclear Information System (INIS)

Es-Saheb, M.H.H.

1990-01-01

The condition-based maintenance of the nuclear reactor, by monitoring and measuring the instantaneous dynamic pressure distribution of the coolant (water) impact on the solid surfaces of the reactor during operation is presented. The behaviour of water domes (jets) produced by underwater explosions of small changes of P.E.T.N. at various depths in two different size cylindrical containers, which simulate the nuclear reactor, is investigated. Water surface domes (jets) from the underwater explosions are photographed. Depending on the depth of the charge, curved and flat top jets of up to 455 mm diameter and impact speeds of up to 70 m/sec. are observed. The instabilities in the dome surfaces are observed and the instantaneous profiles are analysed. It is found that, in all cases tested, the maximum pressure takes place at the center of the jet and could reach up to 3.0 times the on-dimensional impact pressure value. The use of their measurements, as online monitoring for condition-based maintenance and design-out maintenance is discussed. 18 refs

Kartini reactor tank inspection using NDT method for safety improvement of the reactor operation

International Nuclear Information System (INIS)

Syarip; Sutondo, Tegas; Saleh, Chaerul; Nitiswati; Puradwi; Andryansah; Mudiharjo

2002-01-01

The inspection of Kartini reactor tank liner (TRK) by using Non Destructive Testing (NDT) methods to improve the reactor operation safety, have been done. The type of NDT used were: visual examination using an underwater camera and magnifier, replication survey using dental putty, hardness test using an Equotip D indentor, thickness test using ultrasonic probe, and dye penetrant test. The visual examination showed that the surface of TRK was in good condition. The hardness readings were considered to be consistent with the original condition of the tank and the slight hardness increase at the reactor core area consistent with the neutron fluence experienced -10 1 4 n/cm 2 . Results of ultrasonic thickness survey showed that in average the TRK thickness is between 5,0 mm - 6,5 mm, a low 2,1 mm thickness exists at the top of the TRK in the belt area (double layer aluminum plat, therefore do not influencing the safety ). The replica and dye penetrant test at the low thickness area and several suspected areas showed that it could be some defect from original manufacture. Therefore, it can be concluded that the TRK is still feasible for continued operation safely

The Conceptual Design for Tubular Fuel Assemblies of an Advanced Research Reactor

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Dan, Ho Jin; Cho, Yeong Garp; Yoon, Doo Byung; Park, Cheol

2005-05-01

An Advanced Research Reactor(ARR) is being designed by KAERI since 2002. The final goal of the project is to develop a new and unique research reactor model which is superior in safety and economical aspects. In this work, the conceptual design for tubular fuel assemblies was carried out to enhance the previous model. The shape optimization of the cross section of the top guide was performed, and the swaging procedure in connecting fuel plates and stiffeners was developed. Moreover to reflect changes in number and size of fuel plates, related parts of the standard and the reduced fuel assemblies were redesigned. The top guide should suppress the vibration of the fuel assembly due to coolant and resist against material failures owing to fatigue and yield. In order to gain these design requirements, we have optimized the section profile of the top guide. To confirm manufacturing aspects, the swaging procedure was developed and its performance was tested. The results of tangential tensile test and axial compression test guaranteed that the fixing state between fuel plates and stiffeners is firm enough to hold each other. In addition, due to changes in number and size of fuel plates, the outer cross section of the fuel assembly was expanded and the diameter of the spacer tube was reduced. Reflecting these design changes, top/bottom guide, top guide cover, spring, spring cover, and receptacle were readjusted. Based on the technical experiences on the design and operation of the HANARO, the standard and the reduced fuel assemblies will be verified by performing various tests and analysis

Design and Construction of Pool Door for Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Jung, Kwangsub; Lee, Sangjin; Choi, Jinbok; Oh, Jinho; Lee, Jongmin [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The pool door is a structure to isolate the reactor pool from the service pool for maintenance. The pool door is installed before the reactor pool is drained. The pool door consists of structural component and sealing component. The main structures of the pool door are stainless steel plates and side frames. The plates and frames are assembled by welded joints. Lug is welded at the top of the plate. The pool door is submerged in the pool water when it is used. Materials of the pool door should be resistive to corrosion and radiation. Stainless steel is used in structural components and air nozzle assemblies. Features of design and construction of the pool door for the research reactor are introduced. The pool door is designed to isolate the reactor pool for maintenance. Structural analysis is performed to evaluate the structural integrity during earthquake. Tests and inspections are also carried out during construction to identify the safety and function of the pool door.

Design and Construction of Pool Door for Research Reactor

International Nuclear Information System (INIS)

Jung, Kwangsub; Lee, Sangjin; Choi, Jinbok; Oh, Jinho; Lee, Jongmin

2016-01-01

The pool door is a structure to isolate the reactor pool from the service pool for maintenance. The pool door is installed before the reactor pool is drained. The pool door consists of structural component and sealing component. The main structures of the pool door are stainless steel plates and side frames. The plates and frames are assembled by welded joints. Lug is welded at the top of the plate. The pool door is submerged in the pool water when it is used. Materials of the pool door should be resistive to corrosion and radiation. Stainless steel is used in structural components and air nozzle assemblies. Features of design and construction of the pool door for the research reactor are introduced. The pool door is designed to isolate the reactor pool for maintenance. Structural analysis is performed to evaluate the structural integrity during earthquake. Tests and inspections are also carried out during construction to identify the safety and function of the pool door

Loss of Coolant Accident Simulation for the Top-Slot break at Cold Leg Focusing on the Loop Seal Reformation under Long Term Cooling with the ATLAS

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Rok; Park, Yu Sun; Bae, Byoung Uhn; Choi, Nam Hyun; Kang, Kyoung Ho; Choi, Ki Yong [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In the present paper, loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS, which is a thermal-hydraulic integral effect test facility for evolutionary pressurized water reactors (PWRs) of an advanced power reactor of 1400 MWe (APR1400). The simulation was focused on the loop seal reformation under long term cooling condition. During a certain class of Loss of Coolant Accident (LOCA) in a PWR like an advanced power reactor of 1400 MWe (APR1400), the steam volume in the reactor vessel upper plenum and/or upper head may continue expanding until steam blows liquid out of the intermediate leg (U-shaped pump suction cold leg), called loop seal clearing (LSC), opening a path for the steam to be relieved from the break. Prediction of the LSC phenomena is difficult because they are varies for many parameters, which are break location, type, size, etc. This LSC is the major factor that affects the coolant inventory in the small break LOCA (SBLOCA) or intermediate break LOCA (IBLOCA). There is an issue about the loop seal reformation that liquid refills intermediate leg and blocks the steam path after LSC. During the SBLOCA or IBLOCA, the Emergency Core Cooling System (ECCS) is operated. For long term of the top slot small or intermediate break at cold leg, the primary steam condensation by SG heat transfer or SIP, SIT water flooding (reverse flow to loop seal) make loop seal reformation possibly. The primary pressure increase at the top core region due to the steam release blockage by loop seal reformation. And then core level decreases and partial core uncover may occur. The loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS. The loop seal clearing and loop seal reformation were occurred repeatedly.

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1984-12-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

Naisjuhtidega ettevõtete TOP 100

Index Scriptorium Estoniae

2004-01-01

Naisjuhtidega ettevõtete TOP 100. Käibe TOP 20. Käibe kasvu TOP 20. Kasumi TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Riigi- ja kohaliku omavalitsuse asutuste naisjuhtide TOP 25. Riigi- ja kohaliku omavalitsuse asutuste eelarve TOP 25. Riigi- ja kohaliku omavalitsuse asutuste töötajate arvu TOP 25. Riigi- ja kohaliku omavalitsuse asutuste naisjuhtide palga TOP 25

CDF Top Physics

Science.gov (United States)

Tartarelli, G. F.; CDF Collaboration

1996-05-01

The authors present the latest results about top physics obtained by the CDF experiment at the Fermilab Tevatron collider. The data sample used for these analysis (about 110 pb{sup{minus}1}) represents almost the entire statistics collected by CDF during four years (1992--95) of data taking. This large data size has allowed detailed studies of top production and decay properties. The results discussed here include the determination of the top quark mass, the measurement of the production cross section, the study of the kinematics of the top events and a look at top decays.

Naisjuhtidega ettevõtete TOP 70

Index Scriptorium Estoniae

2003-01-01

Naisjuhtidega ettevõtete TOP 70. Käibe TOP 30. Käibe kasvu TOP 30. Kasumi TOP 30. Kasumi kasvu TOP 30. Rentaabluse TOP 30. Omakapitali tootlikkuse TOP 30. Nais- ja meesjuhtidega ettevõtted 2001. aasta 500 edukama ettevõtte hulgas. Naistegevjuhtidega firmade TOP 2001 ja 2002. Riigi- ja kohaliku omavalitsuse asutuste naisjuhtide TOP 20. Riigi- ja kohaliku omavalitsuse asutuste eelarve TOP 20. Riigi- ja kohaliku omavalitsuse asutuste töötajate arvu TOP 20. Naistegevjuhtidega firmade osakaal 2001. aastal. Riigi- ja kohaliku omavalitsuse asutuste naisjuhtide aastapalga TOP 20. Kommenteerib Tiina Raitviir.

Review of ageing management of NPPs - Experience feed back form research reactors

International Nuclear Information System (INIS)

Bhatnagar, A.; Gujarathi, R.I.; Chowdhury, R.; Tikku, A.C.

2002-01-01

Ageing of Systems, Structures and Components (SSCs) is a natural process and sets in along with the construction and commissioning of plants in spite of best design provisions and maintenance practices. Plant operators and maintainers need to plan and take measures against ageing degradation of SSCs to maintain the high standards of safety. As safety is a continuously evolving phenomenon, incorporating safety upgrades from time to time and carrying out ageing management towards improved safety for research and power reactors is very important. Cirus research reactor which was commissioned in 1960 and Tarapur Atomic power station which was commissioned in 1969 are two such examples of older generation nuclear plants in India which are presently undergoing extensive refurbishment towards implementation of ageing management programme. The 40 MWt Cirus Research Reactor located at the Bhabha Atomic Research Centre, Mumbai, is a vertical closed tank type reactor with natural uranium as fuel, demineralised light water as primary coolant, heavy water as moderator and graphite as reflector. The reflector and the thermal shields are cooled by reactor building ventilation system. Sea water is used as secondary coolant. The reactor vessel is made of aluminium and has 199 lattice tubes rolled into top and bottom tube sheets. It has an expansion joint between the top tube sheet and the shell to allow for thermal expansion. The reactor operated very efficiently till early nineties after which the ageing degradation of SSCs started affecting the reactor operation. Plant availability factor showed a declining trend due to frequent breakdown of equipment. Detailed performance review was carried out for various equipment and a list of equipment that needed replacement was prepared. Equipment, for which availability of spares was becoming difficult due to obsolescence, were also included in this list. Detailed ageing studies were then taken up on various SSCs. The SSCs were

Audiitorfirmade TOP 50 aastal 2000

Index Scriptorium Estoniae

2001-01-01

Audiitorfirmade käibe TOP 50, käibe kasvu TOP 25, käibe languse TOP 15, kasumi TOP 50, kasumi kasvu TOP 10, kasumi languse TOP 10, audiitorfirmade finantsnäitajad. Rentaabluse TOP 50, varade tootlikkuse TOP 50

Core disruptive accident analysis in prototype fast breeder reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Kannan, S.E.; Singh, Om Pal; Chetal, S.C.; Bhoje, S.B.

2002-01-01

Liquid metal cooled fast breeder reactors, in particular, pool type have many inherent and engineered safety features and hence a core disruptive accident (CDA) involving melt down of the whole core is a very low probable event ( -6 /ry). The important mechanical consequences such as straining of the main vessel including top shield, structural integrity of safety grade decay heat exchangers (DHX) and intermediate heat exchangers (IHX) sodium release to reactor containment building (RCB) through the penetrations in the top shield, sodium fire and consequent temperature and pressure rise in RCB are theoretically analysed using computer codes. Through the analyses with these codes, it is demonstrated that an energetic CDA capability to the maximum 100 MJ mechanical energy in PFBR can be well contained in the primary containment. The sodium release to RCB is 350 kg and pressure rise in RCB is ∼10 kPa. In order to raise the confidence on the theoretical predictions, very systematic experimental program has been carried out. Totally 67 tests were conducted. This experimental study indicated that the primary containment is integral. The main vessel can withstand the energy release of ∼1200 MJ. The structural integrity of IHX and DHX is assured up to 200 MJ. The transient force transmitted to reactor vault is negligible. The average water leak measured under simulated tests for 122 MJ work potential is about 1.8 kg and the maximum leak is 2.41 kg. Extrapolation of the measured maximum leak based on simulation principles yields ∼ 233 kg of sodium leak in the reactor. Based on the above-mentioned theoretical and experimental investigations, the design pressure of 20 kPa is used for PFBR

COMSORS: A light water reactor chemical core catcher

International Nuclear Information System (INIS)

Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.

1997-01-01

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B 2 O 3 ) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

Support structure for reactor core constituent element

International Nuclear Information System (INIS)

Aida, Yasuhiko.

1993-01-01

A connection pipe having an entrance nozzle inserted therein as a reactor core constituent element is protruded above the upper surface of a reactor core support plate. A through hole is disposed to the protruding portion of the connection pipe. The through hole and a through hole disposed to the reactor core support plate are connected by a communication pipe. A shear rod is disposed in a horizontal portion at the inside of the communication pipe and is supported by a spring horizontally movably. Further, a groove is disposed at a position of the entrance nozzle opposing to the shear rod. The shear rod is urged out of the communication pipe by the pressure of the high pressure plenum and the top end portion of the shear rod is inserted to the groove of the entrance nozzle during operation. Accordingly, the shear rod is positioned in a state where it is extended from the through hole of the communication pipe to the groove of the entrance nozzle. This can mechanically constrain the rising of the reactor core constituent elements by the shear rod upon occurrence of earthquakes. (I.N.)

Jaekaubanduse TOP 100 aastal 2001

Index Scriptorium Estoniae

2002-01-01

TOP 100. Käibe TOP 30. Käibe kasvu TOP 30. Kasumi TOP 30. Kasumi kasvu TOP 30. Rentaabluse TOP 30. Varade tootlikkuse TOP 30. Jaekaubandusettevõtete finantsseadmed. Jaekaubandusettevõtete üldandmed

Arvutifirmade TOP 101 aastal 2004

Index Scriptorium Estoniae

2005-01-01

Arvutifirmade TOP 101; Käibe TOP 20; Käibe kasvu TOP 15; Kasumi TOP 15; Rentaabluse TOP 20; Kasumi kasvu TOP 15; Omakapiali tootlikkuse TOP 15; Eesti arvutifirmade finantsandmed; Arvutifirmade üldandmed

Majutusasutuste TOP 40 aastal 2002

Index Scriptorium Estoniae

2003-01-01

Majutusasutuste TOP 40 aastal 2002. Käibe TOP 40. Kasumi TOP 40. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Majutusasutuste üldandmed. Majutusasutuste finantsandmed

Põlvamaa ettevõtete TOP 50

Index Scriptorium Estoniae

2004-01-01

Ettevõtete TOP 50. Käibe TOP 40. Kasumi TOP 40. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Põlvamaa firmade üldandmed. Põlvamaa firmade finantsandmed

Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

2010-01-01

Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

Eesti Ettevõtete TOP 100

Index Scriptorium Estoniae

2002-01-01

TOP 100. Käibe TOP 500. Käibe kasvu TOP 100. Kasumi TOP 100. Kasumi kasvu TOP 100. Rentaabluse TOP 100. Omakapitali tootlikkuse TOP 100. Eesti edukamate ettevõtete üldandmed. Eesti edukamate ettevõtete finantsnäitajad. Valdkonna ja maakonna TOP-ide edukamate ettevõtete finantsnäitajad

Value addition initiatives for CANDU reactor operation performance

International Nuclear Information System (INIS)

Chugh, V.; Parmar, R.; Schut, J.; Sherin, J.; Xie, H.; Zobin, D.

2013-01-01

Recently, AMEC NSS initiated projects for CANDU® station performance engineering with potentially high returns for the utilities. This paper discusses three initiatives. Firstly, optimization of instrument calibration interval from 1 to 3 years will reduce time commitments on the maintenance resources on top of financial savings ~$3,500 per instrument. Secondly, reactor thermal power uncertainty assessment shows the level of operation which is believed to have an over-conservative margin that can be used to increase power by up to 0.75%. Finally, as an alternative means for controlling Reactor Inlet Header Temperature (RIHT), physical modifications to the High Pressure (HP) feedwater heaters can be useful for partially recovering RIHT resulting in increased production by 10-12 MWe. (author)

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C.

1996-01-01

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses ∼80 W(electric)

Viljandimaa ettevõtete TOP 50

Index Scriptorium Estoniae

2005-01-01

Viljandimaa ettevõtete TOP 50; Käibe TOP 35; Kasumi TOP 35; Käibe kasvu TOP 20; Kasumi kasvu TOP 20; Rentaabluse TOP 20; Omakapitali tootlikkuse TOP 20; Viljandimaa ettevõtete üld- ja finantsandmed

Jõgevamaa ettevõtete TOP 50

Index Scriptorium Estoniae

2005-01-01

Jõgevamaa ettevõtete TOP 50; Käibe TOP 35; Kasumi TOP 35; Käibe kasvu TOP 20; Kasumi kasvu TOP 20; Rentaabluse TOP 20; Omakapitali tootlikkuse TOP 20; Jõgevamaa ettevõtete üld- ja finantsandmed

Jõgevamaa ettevõtete TOP 55

Index Scriptorium Estoniae

2004-01-01

Jõgevamaa ettevõtete TOP 55 aastal 2003. Käibe TOP 40. Kasumi TOP 40. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Jõgevamaa firmade üld- ja finantsandmed

Turismifirmade 2000. a. TOP 30

Index Scriptorium Estoniae

2001-01-01

Turismiettevõtete üldandmed: turismiettevõtete finantsnäitajad; käibe TOP 30; käibe kasvu TOP 10; kasumi TOP 20; kasumi kasvu TOP 10; kasumi languse TOP; rentaabluse TOP 10; varade tootlikkuse TOP 10

Online-kaubanduse TOP aastal 2003

Index Scriptorium Estoniae

2004-01-01

Online-kaubanduse TOP aastal 2003. Käibe TOP 5. Käibe kasvu TOP 5. Rentaabluse TOP 5. Kasumi TOP 5. Kasumi kasvu TOP 5. Omakapitali tootlikkuse TOP 5. Lisa: TOPi koostamise metoodika. Online-kaubanduse firmade üldandmed. Online kaubanduse firmade finantsandmed

Ehitusmaterjalitootjate TOP 70 aastal 2003

Index Scriptorium Estoniae

2004-01-01

Ilmunud ka: Delovõje Vedomosti : Stroitelstvo, 29. sept. 2004, lk. 2,4. Ehitusmaterjalitootjate TOP 70; Käibe TOP 10; Käibe kasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP. Ehitusmaterjalitootjate üldandmed

The success of operation and utilization of the Indonesia multipurpose reactor G.A. Siwabessy

International Nuclear Information System (INIS)

Taryo, Taswanda; Kuntoro, Iman

2000-01-01

The Indonesia Multipurpose Reactor G.A. Siwabessy (RSG-GAS), operated by Multipurpose Reactor Center (MPRC/PRSG-BATAN), went its first criticality in July 1987. The reactor then achieved the power of 30 MW thermal in March 1992. Based on user requirement, the reactor is usually operated at the power of 20 MW thermal. The RSG-GAS is put to use mainly for radioisotope production, R and D on reactor safety and by using beam tubes, the reactor can also be applied for R and D on science and materials. Operation and maintenance of the reactor have been well organized due to well technical and administrative management from the top manager to all people involved in those two activities. Within their support, the RSG-GAS has occupied great advantages not only for man power development in our center but also for scientific cooperation with whoever would like to apply the RSG-GAS for R and D with mutual benefit agreement. (author)

Microbial Aggregate and Functional Community Distribution in a Sequencing Batch Reactor with Anammox Granules

KAUST Repository

Sun, Shan

2013-05-01

Anammox (anaerobic ammonium oxidation) process is a one-step conversion of ammonia into nitrogen gas with nitrite as an electron acceptor. It has been developed as a sustainable technology for ammonia removal from wastewater in the last decade. For wastewater treatment, anammox biomass was widely developed as microbial aggregate where the conditions for enrichment of anammox community must be delicately controlled and growth of other bacteria especially NOB should be suppressed to enhance nitrogen removal efficiency. Little is known about the distribution of microbial aggregates in anammox process. Thus the objective of our study was to assess whether segregation of biomass occurs in granular anammox system. In this study, a laboratory-scale sequential batch reactor (SBR) was successfully operated for a period of 80 days with granular anammox biomass. Temporal and spatial distribution of microbial aggregates was studied by particle characterization system and the distribution of functional microbial communities was studied with qPCR and 16s rRNA amplicon pyrosequencing. Our study revealed the spatial and temporal distribution of biomass aggregates based on their sizes and density. Granules (>200 μm) preferentially accumulated in the bottom of the reactor while floccules (30-200 μm) were relatively rich at the top layer. The average density of aggregate was higher at the bottom than the density of those at the top layer. Degranulation caused by lack of hydrodynamic shear force in the top layer was considered responsible for this phenomenon. NOB was relatively rich in the top layer while percentage of anammox population was higher at the bottom, and anammox bacteria population gradually increased over a period of time. NOB growth was supposed to be associated with the increase of floccules based on the concurrent occurrence. Thus, segregation of biomass can be utilized to develop an effective strategy to enrich anammox and wash out NOB by shortening the settling

Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

International Nuclear Information System (INIS)

Ergun, E; Kadiroglu, O.S.

1999-01-01

The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

Status of the DOE's foreign research reactor spent nuclear fuel acceptance program

International Nuclear Information System (INIS)

Chacey, K.; Saris, E.C.

1997-01-01

In May 1996, the U.S. Department of Energy (DOE), in consultation with the U.S. Department of State (DOS), adopted a policy to accept and manage in the United States ∼20 tonnes of spent nuclear fuel from research reactors in up to 41 countries. This spent fuel is being accepted under the nuclear weapons non-proliferation policy concerning foreign research reactor spent nuclear fuel. Only spent fuel containing uranium enriched in the United States is covered under this policy. Implementing this policy is a top priority of the DOE. The purpose of this paper is to provide the current status of the foreign research reactor acceptance program, including achievements to date and future challenges

Probabilistic safety analysis for the Triga reactor Vienna

International Nuclear Information System (INIS)

Boeck, H.; Kirchsteiger, C.

1988-07-01

Triga-type reactors are the most widely used low power research reactors with power levels up to 3 MW. Although Triga reactors are considered inherently safe, due to their unique features such as prompt negative temperature coefficient and low power density, the reactor core still contains a respectable amount of activity which could lead under very adverse circumstances to radiation exposure both of staff members and of public. Such circumstances could be external events, accidents during fuel element manipulation or a loss of coolant water with exposure of the core. Therefore, it was decided to look more closely to various accident pathways and to calculate their probability, if possible. A major drawback is the lack of statistical material because no centralized registration of failures is carried out. Therefore, in many cases values from other research reactor types or even from power reactor statistics had to be used, thus increasing the uncertainty of the results. As most undesired event or TOP-event in this analysis a radiation exposure of staff members, the public or both together was selected and the probabilities of different pathways leading to this exposure was calculated. In the present case 'radiation exposure' are dose rates or activity concentration above the international accepted limits for occupational staff or public. 20 refs., 10 figs. (Author)

Hiiumaa ettevõtete TOP 50

Index Scriptorium Estoniae

2005-01-01

Hiiumaa ettevõtete TOP 50; Käibe TOP 35; Kasumi TOP 35; Hiiumaa ettevõtete üld- ja finantsandmed; Käibe kasvu TOP 20; Kasumi kasvu TOP 20; Rentaabluse TOP 20; Omakapitali tootlikkuse TOP 20. Vt samas: Teeli Remmalg: Hiiumaal jätkub plastitööstuse võidumarss

Structural mechanics studies at E.D.F

International Nuclear Information System (INIS)

Baylac, G.

1983-01-01

Structural mechanics studies at EDF have three goals: a better knowledge of the materials properties, an an improvement of the design, and a better in service surveillance of the components. This study has lead EDF to perform a large investment to make possible the fatigue survey of the primary circuit. This investment is of 10 men-years for the mechanical studies. A cumulative bookkeeping of the transients is now in action at Fessenheim I and II, Bugey II to V, Tricastin I, Gravelines I, Dampierre I. A catalog of the transients easy to use will be provided to each unit in the near future. The design of the new four loop plants will take advantage of a new catalog of the design transients, this catalog being used for the purpose of the design and the bookkeeping of the transients. Experimental and theoretical investigations concerning the vibrations of PWR internals and primary circuit have been carried out at Fessenheim I, Bugey V and Tricastin I . As a result of these studies and complementary studies on Safran mock-up, EDF has been able to define with FRAMATOME and CEA a monitoring system to meet the requirements of the safety authorities. The monitoring system is divided in to three parts: loose - parts detection system accelerometers; monitoring of reactor internals by neutron noise measurements; monitoring of heavy components vibrations by accelerometers. This system is now installed in all PWR units. Some developments are in progress at EDF mainly at the Directorate of Research and Development to improve the procedures of the control and to define the criteria for an early diagnostic of the anomalies. The major reports are Surveillance du comportement vibratoire des composants de circuit primaire; Vibration studies on a three loop PWR internals model; and Nuclear Reactor Surveillance - Neutron noise measurements and vibrations analysis on French PWR Internal structures

Development of system design and seismic performance evaluation for reactor pool working platform of a research reactor

International Nuclear Information System (INIS)

Kwag, Shinyoung; Lee, Jong-Min; Oh, Jinho; Ryu, Jeong-Soo

2014-01-01

Highlights: • Design of reactor pool working platform (RPWP) is newly proposed for an open-tank-in-pool type research reactor. • Main concept of RPWP is to minimize the pool top radiation level. • Framework for seismic performance evaluation of nuclear SSCs in a deterministic and a probabilistic manner is proposed. • Structural integrity, serviceability, and seismic margin of the RPWP are evaluated during and after seismic events. -- Abstract: The reactor pool working platform (RPWP) has been newly designed for an open-tank-in-pool type research reactor, and its seismic response, structural integrity, serviceability, and seismic margin have been evaluated during and after seismic events in this paper. The main important concept of the RPWP is to minimize the pool top radiation level by physically covering the reactor pool of the open-tank-in-pool type research reactor and suppressing the rise of flow induced by the primary cooling system. It is also to provide easy handling of the irradiated objects under the pool water by providing guide tubes and refueling cover to make the radioisotopes irradiated and protect the reactor structure assembly. For this concept, the new three dimensional design model of the RPWP is established for manufacturing, installation and operation, and the analytical model is developed to analyze the seismic performance. Since it is submerged under and influenced by water, the hydrodynamic effect is taken into account by using the hydrodynamic added mass method. To investigate the dynamic characteristics of the RPWP, a modal analysis of the developed analytical model is performed. To evaluate the structural integrity and serviceability of the RPWP, the response spectrum analysis and response time history analysis have been performed under the static load and the seismic load of a safe shutdown earthquake (SSE). Their stresses are analyzed for the structural integrity. The possibility of an impact between the RPWP and the most

Top-down workforce demand extrapolation based on an EC energy road-map scenario

International Nuclear Information System (INIS)

Roelofs, F.; Von Estorff, U.

2014-01-01

The EHRO-N team of JRC-IET provides the EC with essential data related to supply and demand for nuclear experts based on bottom-up information from the nuclear industry. The current paper deals with an alternative approach to derive figures for the demand side information of the nuclear workforce. Complementary to the bottom-up approach, a top-down modelling approach extrapolation of an EC Energy road-map nuclear energy demand scenario is followed here in addition to the survey information. In this top-down modelling approach, the number of nuclear power plants that are in operation and under construction is derived as a function of time from 2010 up to 2050 assuming that the current reactor park will be replaced by generic third generation reactors of 1400 MWe or 1000 MWe. Depending on the size of new build reactors, the analysis shows the number of new reactors required to fulfil the demand for nuclear energy. Based on workforce models for operation and construction of nuclear power plants, the model allows an extrapolation of these respective work-forces. Using the nuclear skills pyramid, the total workforce employed at a plant is broken down in a nuclear (experts), nuclearized, and nuclear aware workforce. With retirement profiles for nuclear power plants derived from the bottom-up EHRO-N survey, the replacement of the current workforce is taken into account. The peak of the new workforce (partly replacing the retiring workforce and additionally keeping up with the growing total workforce demand) for nuclear experts and nuclearized employees is to be expected at the end of the considered period (2050). However, the peak workforce for nuclear aware employees is to be expected around 2020. When comparing to historical data for the nuclear capacity being installed at the same time in Europe, it is clear that the expected future capacity to be installed at the same time in Europe is significantly lower (factor of 2) than in the early 1980's. However, it should

Design and development of face seal type sealing plug for advanced heavy water reactor

International Nuclear Information System (INIS)

Bansal, S.; Bhattacharyya, S.; Patel, R.J.; Agrawal, R.G.; Vaze, K.K.

2005-09-01

Advanced Heavy Water Reactor is a vertical pressure tube type reactor having light water as its coolant and heavy water as moderator. Sealing plug is required to close the pressure boundary of main heat transport system of the reactor by preventing escape of light water/steam From the coolant channel. There are 452 coolant channels in the reactor located in square lattice pitch. Sealing plug is located at the top of each coolant channel (in the top end fitting). Top end fitting is having a stepped bore to create a sealing face. Sealing plug is held through its expanded jaws in a specially provided groove of the end fitting. The plug was designed and prototypes were manufactured considering its functional importance, intricate design and precision machining requirements. Sealing plug consists of about 20 components mostly made up of precipitation hardening stainless steel, which is suitable for water environment and meets other requirements of strength and resistance to wear and galling. Seal disc is a critical component of the sealing plug as it is the pressure-retaining component. It is a circular disc with protruded stem. One face of the seal disc is nickel plated in the peripheral area that creates the sealing by abutting against the sealing face provided in the end fitting. The typical shape and profile of seal disc provides flexibility and allows elastic deformation to assist in locking of sealing plug and creating adequate seating force for effective sealing. Design and development aspects of the sealing plug have been detailed out in this report. Also results of stress analysis and experimental studies for seal disc have been mentioned in the report. Stress analysis and experimental testing was required for the seal disc because high stresses are developed due to its exposure to high pressure and temperature environment of Main Heat Transport system. Hot testing was carried out to simulate the reactor-simulated condition. The performance was found to be

Current status of work on preservation of accumulated knowledge on fast reactors in Russia and plan of top-priority measures

International Nuclear Information System (INIS)

Kotchetkov, L.A.; Poplavsky, V.M.; Tsiboulya, A.M.; Ashurko, Yu.M.

2005-01-01

The future of nuclear power is associated with mastering of fast reactor technology. Experience gained in Russia by now in the development of sodium cooled fast reactors (FR) and reactor plants of various applications is one of the most extensive and successful all over the world. Since the late 1940-ies up to now, well-directed, rather intensive work has been carried out in the USSR (later in Russia) on all aspects of fast reactors. Institute for Physics and Power Engineering has been always leading organization in the USSR and in Russia in the area of fast reactors. Work on fast reactors in Russia was carried out by many institutions, namely: IPPE, VNIINM, OKBM, VNIPIET, OKB Gidropress, RIAR, SPbAEP, TsNII Prometey and teams of the BN-350 and the BN-600 plants working in close and fruitful cooperation. Successful operation of the power unit of Beloyarskaya NPP with the BN-600 fast reactor during 25 years is one of the good results of this vast expensive efforts. In view of delay in wide-scale deployment of fast neutron reactors and change of generations of specialists in the area of FR, a necessity has arisen in the preservation of knowledge and experience on FR gained in many countries including Russia. Certain measures in this area have been planned by the Russian organizations. However, the necessity has become imminent in a purposeful systematized approach to the preservation of knowledge in fast reactor area, which can be realized only within the framework of development of appropriate special work program. The basic work trends within the framework of this program have been stated. In view of urgency of some part of this work, it is necessary to prioritize the work contents. IAEA assistance (methodological, organizational and financial) in the implementation of some sections of the program would facilitate successful implementation of the work program on preservation of knowledge on FR in Russia. (author)

Reactor fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.; Groves, M.D.

1980-01-01

A nuclear reactor fuel assembly having a lower end fitting and actuating means interacting therewith for holding the assembly down on the core support stand against the upward flow of coolant. Locking means for interacting with projections on the support stand are carried by the lower end fitting and are actuated by the movement of an actuating rod operated from above the top of the assembly. In one embodiment of the invention the downward movement of the actuating rod forces a latched spring to move outward into locking engagement with a shoulder on the support stand projections. In another embodiment, the actuating rod is rotated to effect the locking between the end fitting and the projection. (author)

Study on the application of CANDLE burnup strategy to several nuclear reactors. JAERI's nuclear research promotion program, H13-002 (Contract research)

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko

2005-03-01

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. When this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means that the nuclear energy can be utilized for many hundreds years without new mining, enrichment and reprocessing, and the amount of spent fuel can be reduced considerably. However, in order to perform such a high fuel burnup some innovative technologies should be developed. Though development of innovative fuel will take a lot of time, intermediate re-cladding may be easy to be employed. Compared to fast reactors, application of CANDLE burnup to prismatic fuel high-temperature gas cooled reactors is very easy. In this report the application of CANDLE burnup to both these types of reactors are studied. (author)

On economic efficiency of nuclear power unit life extension using steam-gas topping plant

International Nuclear Information System (INIS)

Kuznetsov, Y.N.; Lisitsa, F.D.; Smirnov, V.G.

2001-01-01

The different options for life extension of the operating nuclear power units have been analyzed in the report with regard for their economic efficiency. A particular attention is given to the option envisaging the reduction of reactor power output and its subsequent compensation with a steam-gas topping plant. Steam generated at its heat-recovery boilers is proposed to be used for the additional loading of the nuclear plant turbine so as to reach its nominal output. It would be demonstrated that the implementation of this option allows to reduce total costs in the period of power plant life extension by 24-29% as compared with the alternative use of the replacing steam-gas unit and the saved resources could be directed, for instance, for decommissioning of a reactor facility. (authors)

Vessel core seismic interaction for a fast reactor

International Nuclear Information System (INIS)

Martelli, A.; Maresca, G.

1984-01-01

This report deals with the analysis carried out in collaboration between ENEA and NIRA for optimizing the iterative procedure applied for the evaluation of the effects of the vessel core dynamic interaction for a fast reactor in the case of a earthquake. In fact, as shown in a previous report the convergence of such procedure was very slow for the design solution adopted for the PEC reactor, i.e. with a core restraint plate located close to the top of the core elements. This study, although performed making use of preliminary data (the same of the cited previous report) demonstrates that the convergence is fast if a suitable linear core model is applied in the first iteration linear calculations carried out by NIRA, with an intermediate stiffness with respect to those corresponding to the two limit models previously assumed and increased damping coefficients. Thus, the optimized iterative procedures is now applied in the PEC reactor block seismic verification analysis

Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

Energy Technology Data Exchange (ETDEWEB)

Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2011-07-01

The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

Audiitorite TOP 50 aastal 2005

Index Scriptorium Estoniae

2006-01-01

Audiitorite TOP. Vt. samas: Käibe TOP 10; Käibe kasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; ROE TOP 10; Ketlin Priilinn. Kasvu tagavad lojaalsed kliendid; Klient on audiitori parim müügimees; Teeli Remmelg. Kliendid vaatavad pigem kvaliteeti kui madalat hinda. Kommenteerivad Signe Keernik ja Kalle Lahe. Tabel: Audiitoriettevõtete üld- ja finantsandmed

Enhanced Westinghouse WWER-1000 fuel design for Ukraine reactors

International Nuclear Information System (INIS)

Dye, M.; Shah, H.

2015-01-01

Westinghouse has completed design, development, and region quantity delivery of an enhanced Westinghouse fuel assembly for WWER-1000 reactors to support continued safe reactor operations. The enhanced design builds on the successful performance of an earlier generation design which has operated in the South Ukraine 3 reactor for multiple cycles without any fuel rod failures. Incorporated design enhancements include a thicker spacer grid outer strap, an enhanced spacer grid outer strap profile to limit the risk for, and impact of, mechanical interaction/interference with coresident fuel, an all Alloy 718 grid structure for improved stability and strength, and improvements to the top and bottom nozzles. Capable of meeting increased lateral loads generated from using a higher axial trip limit for the refueling machine crane, the design was verified by extensive mechanical and thermalhydraulic testing, which included a newly developed fuel assembly-to-fuel assembly handling test rig to assess performance during bounding core loading and unloading conditions. Through these extensive design enhancements and comprehensive testing program, the enhanced WWER-1000 design provides additional performance, handling, and reliability margins for safe reactor operation. (authors)

At the heart of EDF's education plant

International Nuclear Information System (INIS)

Maillard, C.

2010-01-01

Facing a high and growing number of retirements and the present revival of nuclear energy, Electricite de France (EDF), the French state-owned electric utility, is struggling to find the competencies it needs, especially in domains such as reactor operators, mechanic technicians and engineers in the fields of fittings, boiler fabrication, maintenance, automation, etc. EDF is thus engaged in an important education and training program. The author reports from Bugey (France), where is located one of the main EDF production-engineering training center, and describes the variety of learning activities available in the center. An important effort is devoted to the training of foreign operators in order for EDF to better compete on the nuclear energy international market

Intergenerational Top Income Persistence

DEFF Research Database (Denmark)

Munk, Martin D.; Bonke, Jens; Hussain, M. Azhar

2016-01-01

In this paper, we investigate intergenerational top earnings and top income mobility in Denmark. Access to administrative registers allowed us to look at very small fractions of the population. We find that intergenerational mobility is lower in the top when including capital income in the income...... measure— for the rich top 0.1% fathers and sons the elasticity is 0.466. Compared with Sweden, however, the intergenerational top income persistence is about half the size in Denmark....

FBR type reactor

International Nuclear Information System (INIS)

Nagai, Fumio.

1979-01-01

Purpose: To unify the temperature distribution in a nuclear reactor vessel by the provision of a gas recycle path for pressurizing a cover gas to recycle the cover gas and thus stir the gas in a cover gas chamber. Constitution: A plurality of gas inlet tubes and gas discharge tubes are provided to the wall of a cover gas chamber above the liquid level of coolants in a nuclear reactor vessel and the cover gas is recycled through the tubes. The plurality of gas inlet tubes are each provided at their tops with nozzles opening circumferentially and communicated to the outlet of a compressor. While on the other hand, the plurality of gas discharge tubes are communicated to the inlet of a compressor. Upon operation of the compressor, the pressurized cover gas is jetted out from the nozzles, swirls along the inner circumferential surface of the vessel and interrupts and stirs the vertical thermal convection. The gas, after swirling one-half of the inner circumferential surface of the vessel, automatically flows out of the gas discharging tubes opening behind the nozzles and then flows into the inlet of the compressor. (Seki, T.)

Reactor shutdown device

International Nuclear Information System (INIS)

Ito, Masahiko

1990-01-01

The object of the present invention is to reliably shutdown an LMFBR type reactor upon accident of the reactor. That is, curie point magnetic member is made annular so that it can be moved between the outer circumference of an electromagnet and the position above the electromagnet. This enables to enlarge the curie point magnetic member since it is no more necessary to be inserted it in a guide tube. Accordingly, attracting force upon normal operation is increased to remarkably improve the reliability against erronerous scram, etc. Further, since a required gap is formed between the curie point magnetic member and the electromagnet and the heat of coolants is efficiently transmitted to the curie point magnetic member, rapid scram is possible. Further, a position support mechanism is disposed to a part of a control element or at the inner side of the guiding tube for urging and actuating the armature to make it protrude above the top of the guiding tube. With such a constitution, since the armature can be adsorbed without inserting the curie point magnetic member and the electromagnet guide tube, the same effect as in the case of inserting them can be obtained. (I.S.)

Top quark discovered

International Nuclear Information System (INIS)

Anon.

1995-01-01

Nine months after a careful announcement of tentative evidence for the long-awaited sixth 'top' quark, physicists from the CDF and DO experiments at Fermilab's Tevatron proton-antiproton collider declared on 2 March that they had finally discovered the top quark. Last year (June 1994, page 1), the CDF experiment at the Tevatron reported a dozen candidate top events. These, said CDF, had all the characteristics expected of top, but the difficulties of extracting the tiny signal from a trillion proton-antiproton collisions made them shy of claiming a discovery. For its part, the companion DO Tevatron experiment reported a few similar events but were even more guarded about their interpretation as top quarks. Just after these hesitant announcements, performance at the Tevatron improved dramatically last summer. After the commissioning of a new linear accelerator and a magnet realignment, the machine reached a new world record proton-antiproton collision luminosity of 1.28 x 10 31 per sq cm per s, ten times that originally planned. Data began to pour in at an unprecedented rate and the data sample grew to six trillion collisions. Luminosity has subsequently climbed to 1.7 x 10 31 . The top quark is the final letter in the alphabet of Standard Model particles. According to this picture, all matter is composed of six stronglyinteracting subnuclear particles, the quarks, and six weakly interacting particles, the leptons. Both sextets are neatly arranged as three pairs in order of increasing mass. The fifth quark, the 'beauty' or 'b' quark, was also discovered at Fermilab, back in 1977. Since then physicists have been eagerly waiting for the top to turn up, but have been frustrated by its heaviness - the top is some 40 times the mass of its 'beautiful' partner. Not only is the top quark the heaviest by far, but it is the only quark which has been actively hunted. After the quarry was glimpsed last year, the net has now been

Uncovering the single top: observation of electroweak top quark production

Energy Technology Data Exchange (ETDEWEB)

Benitez, Jorge Armando [Michigan State Univ., East Lansing, MI (United States)

2009-01-01

The top quark is generally produced in quark and anti-quark pairs. However, the Standard Model also predicts the production of only one top quark which is mediated by the electroweak interaction, known as 'Single Top'. Single Top quark production is important because it provides a unique and direct way to measure the CKM matrix element V_tb, and can be used to explore physics possibilities beyond the Standard Model predictions. This dissertation presents the results of the observation of Single Top using 2.3 fb^-1 of Data collected with the D0 detector at the Fermilab Tevatron collider. The analysis includes the Single Top muon+jets and electron+jets final states and employs Boosted Decision Tress as a method to separate the signal from the background. The resulting Single Top cross section measurement is: (1) σ(p$\\bar{p}$→ tb + X, tqb + X) = 3.74_-0.74^+0.95 pb, where the errors include both statistical and systematic uncertainties. The probability to measure a cross section at this value or higher in the absence of signal is p = 1.9 x 10^-6. This corresponds to a standard deviation Gaussian equivalence of 4.6. When combining this result with two other analysis methods, the resulting cross section measurement is: (2) σ(p$\\bar{p}$ → tb + X, tqb + X) = 3.94 ± 0.88 pb, and the corresponding measurement significance is 5.0 standard deviations.

Operating experiences of reactor shutdown system at MAPS

International Nuclear Information System (INIS)

Kotteeswaran, T.J.; Subramani, V.A.; Hariharan, K.

1997-01-01

The reactors in Madras Atomic Power Station (MAPS), Kalpakkam are Pressurised Heavy Water Reactors (PHWR) similar to RAPS, Kota. The moderator heavy water is pumped into the calandria from dump tank to make the reactor critical. Later with the calandria level held constant at 92% FT, the further power changes are being done with the movement of adjuster rods. The moderator is held in calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The shutdown of the reactor is effected by dumping the moderator water to dump tank by fast equalizing of helium gas pressure. In the revised mode of operation of moderator circuit after the moderator inlet manifold failure, the dump timing was observed to be more compared to the normal value. This was investigated and observed to be due to accumulation of D 2 O in the gas space above dump valves, which was affecting the helium equalizing flow. Also some of Indicating Alarm Meters (IAM) in protective system initiating the trip signals have failed in the unsafe mode. They have been modified to avoid the recurrence of the failures. (author)

Physics design of advanced steady-state tokamak reactor A-SSTR2

International Nuclear Information System (INIS)

Nishio, Satoshi; Ushigusa, Kenkichi

2000-10-01

Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

Top quark measurements at ATLAS

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00041686; The ATLAS collaboration

2017-01-01

The top quark is the heaviest known fundamental particle. As it is the only quark that decays before it hadronizes, it allows us to probe the properties of bare quarks at the Large Hadron Collider. Highlights of a few recent precision measurements by the ATLAS Collaboration of the top quark using 13 TeV and 8 TeV collision data will be presented: top-quark pair and single top production cross sections including differential distributions will be presented alongside measurements of top-quark properties, including results using boosted top quarks, probe our understanding of top-quark production in the TeV regime. Measurements of the top-quark mass and searches for rare top quark decays are also presented.

Temperature measuring element in nuclear reactors

International Nuclear Information System (INIS)

Wada, Takashi.

1987-01-01

Purpose: To easily measure the partial maximum temperature at a portion within the nuclear reactor where the connection with the external portion is difficult. Constitution: Sodium, potassium or the alloy thereof with high heat expansion coefficient is packed into an elastic vessel having elastic walls contained in a capsule. A piercing member formed into an acute triangle is attached to one end in the direction of expansion and contraction of the elastic container. The two sides of the triangle form an acute knife edge. A diaphragm is disposed within a capsule at a position opposed to the sharpened direction of the piercing member. Such a capsule is placed in a predetermined position of the nuclear reactor. The elastic vessel causes thermal expansion displacement depending on the temperature at a certain position, by which the top end of the pierce member penetrates through the diaphragm. A pierced scar of a penetration length depending on the temperature is resulted to the diaphragm. The length of the piercing damage is electroscopically observed and compared with the calibration curve to recognize the maximum temperature in the predetermined portion of the nuclear reactor. (Kamimura, M.)

Top quark measurements at ATLAS

CERN Document Server

Grancagnolo, Sergio; The ATLAS collaboration

2017-01-01

The top quark is the heaviest known fundamental particle. As it is the only quark that decays before it hadronizes, this gives us the unique opportunity to probe the properties of bare quarks at the Large Hadron Collider. This talk will present highlights of a few recent precision measurements by the ATLAS Collaboration of the top quark using 13 TeV and 8 TeV collision data: top-quark pair and single top production cross sections including differential distributions will be presented alongside top quark properties measurements. These measurements, including results using boosted top quarks, probe our understanding of top quark production in the TeV regime. Measurements of the top quark mass and searches for rare top quark decays are also presented.

Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

International Nuclear Information System (INIS)

Ha, J.J.; Belhadj, M.; Aldemir, T.; Christensen, R.N.

1987-01-01

Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16 N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

Transients in reactors for power systems compensation

Science.gov (United States)

Abdul Hamid, Haziah

surge arrester operation during the MSCDN energisation, which causes steep voltage change at the reactor terminal. (ii) Second, the nonuniform voltage distribution, resulting in high stresses across the top inter-turn windings. (iii) Third, the rapid rate-of-change of voltage in the assumed worst-case reactor winding location. This is accompanied by a high dielectric current through the inter-turn winding insulation..

Tekstiilitööstuse TOP 47

Index Scriptorium Estoniae

2005-01-01

Tekstiilitööstuse TOP 47; Käibe TOP 35; Kasumi TOP 35; Käibe kasvu TOP 20; Kasumi kasvu TOP 20; Rentaabluse TOP 20; Varade tootlikkuse TOP 20; Tekstiilitööstuse TOP-i firmade üld- ja finantsandmed

Rõivatööstuse TOP 50

Index Scriptorium Estoniae

2005-01-01

Rõivatööstuse TOP 50; Käibe TOP 35; Kasumi TOP 35; Käibe kasvu TOP 20; Kasumi kasvu TOP 20; Rentaabluse TOP 20; Omakapitali tootlikkuse TOP 20; Rõivatööstuse TOP-i firmade üld- ja finantsandmed

Top Quark Mass

CERN Document Server

Mulders, Martijn

2016-01-01

Ever since the discovery of the top quark at the Tevatron collider in 1995 the measurement of its mass has been a high priority. As one of the fundamental parameters of the Standard Theory of particle physics, the precise value of the top quark mass together with other inputs provides a test for the self-consistency of the theory, and has consequences for the stability of the Higgs field that permeates the Universe. In this review I will briefly summarize the experimental techniques used at the Tevatron and the LHC experiments throughout the years to measure the top quark mass with ever improving accuracy, and highlight the recent progress in combining all measurements in a single world average combination. As experimental measurements became more precise, the question of their theoretical interpretation has become important. The difficulty of relating the measured quantity to the fundamental top mass parameter has inspired alternative measurement methods that extract the top mass in complementary ways. I wil...

PREFACE: 5th International Workshop on Top Quark Physics (TOP2012)

Science.gov (United States)

Salamanna, G.; Boisvert, V.; Cerrito, L.; Khan, A.; Moretti, S.; Owen, M.; Schwanenberger, C.

2013-07-01

The 5th International Workshop on Top Quark Physics (TOP 2012) took place in Winchester, UK, from the 16-21 September. It gathered students as well as people active in the top quark sector and provided a framework to highlight the newest results and matters related to top quark physics. Discovered in 1995, the top quark is the sixth and heaviest of all quarks, and it is the only one with a lifetime short enough to be observed 'naked'. This makes it an important testing ground in the search for new physics. In fact, the fact of its mass being so much larger than the other quarks, hints at its special role in the Higgs mechanism. For the same reason, in many models of New Physics, new heavy resonances are expected to couple mostly with top quarks. Even if no new particles are observed, the direct correlation between its angular momentum and that of its detectable decay products allows us to probe indirectly New Physics in action when top quarks are created. In this edition of the TOP conference series, for the first time, the agenda was equally balanced between 'traditional' measurements and the now vast number of searches for physics BSM in the top quark sector, thanks mostly to the amount of data collected at the LHC in its Run I. New results were presented by both the Tevatron and the LHC collaborations: improved ttbar and single top cross-section measurements, refined techniques to measure the top quark mass and a large number of results on properties such as spin correlation and W boson polarization in top quark decays were shown. More technical discussions on the experimental issues, both from the detector and the simulation side also took place, drawing together experimentalists and theorists. Reviews of the latest results on ttbar asymmetry both from CDF and D0 and from ATLAS and CMS were shown, and theorists active in the field made some interesting points on this hot topic. Additionally, results on the search for fourth generation fermions and new

Top Quark Physics with CMS

CERN Multimedia

CERN. Geneva

2011-01-01

Higgs mechanism. There are various hints at deviations from the Standard Model expectation which have been observed recently by Tevatron experiments in top final states. Several signatures of new physics accessible at the LHC either suffer from top-quark production as a significant background or contain top quarks themselves. In this talk, we present results on top quark physics obtained from the first LHC data collected by the CMS experiment.They include measurements of the top pair production cross section in various channels and their combination, measurements of the top quark mass, the single top cross section, a search for new particles decaying into top pairs, and a first look at the charge asymmetry.

An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

Energy Technology Data Exchange (ETDEWEB)

Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2009-11-01

This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

Heat load imposed on reactor vessels during in-vessel retention of core melts

Energy Technology Data Exchange (ETDEWEB)

Kim, Su-Hyeon; Chung, Bum-Jin, E-mail: bjchung@khu.ac.kr

2016-11-15

Highlights: • Angular heat load on reactor vessel by natural convection of oxide pool was measured. • High Ra was achieved by using mass transfer experiments based on analogy concept. • Measured Nusselt numbers agreed reasonably with the other existing studies. • Three different types of volumetric heat sources were compared. • They didn’t affect the heat flux of the top plate but affected those of the reactor vessel. - Abstract: We measured the heat load imposed on reactor vessels by natural convection of the oxide pool in severe accidents. Based on the analogy between heat and mass transfer, mass transfer experiments were performed using a copper sulfate electroplating system. A modified Rayleigh number of the order 10{sup 14} was achieved in a small facility with a height of 0.1 m. Three different types of volumetric heat sources were compared and the average Nusselt number of the curved surface was 39% lower, whereas in the case of the top plate was 6% higher than in previous studies with a two-dimensional geometry due to the high Sc value of this study. Reliable experimental data on the angular heat flux ratios were reported compared to those of the BALI and SIGMA CP facilities in terms of fluctuations and consistency.

Top-ophilia

Energy Technology Data Exchange (ETDEWEB)

Quigg, Chris; /Fermilab

2008-01-01

Almost from the moment in June 1977 when the discovery of the Upsilon resonance revealed the existence of what we now call the bottom quark, physicists began searching for its partner. Through the years, as we established the electric charge and weak isospin of the b-quark, and detected the virtual influence of its mate, it became clear that the top quark must exist. Exactly at what mass, we couldn't say, but we knew just how top events would look. We also knew that top events would be rare--if the Tevatron could make them at all--and that picking out the events would pose a real challenge for the experimenters and their detectors.

BWR type reactors

International Nuclear Information System (INIS)

Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

1980-01-01

Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

Toitlustusettevõtete TOP 30 aastal 2002

Index Scriptorium Estoniae

2003-01-01

Toitlustusettevõtete TOP 30 aastal 2002. Käibe TOP 30. Kasumi TOP 30. Käibe kasvu TOP 30. Kasumi kasvu TOP 30. Rentaabluse TOP 30. Omakapitali tootlikkuse TOP 30. Toitlustusettevõtete üldandmed. Toitlustusettevõtete finantsandmed

Toiduainetööstuste TOP 100

Index Scriptorium Estoniae

2006-01-01

Toiduainetööstuse TOP. Vt. samas: Käibe TOP 10; Käibekasvu TOP 10; Kasumi TOP 10; Kasumi kasvu TOP 10; Rentaabluse TOP 10; Omakapitali tootlikkuse TOP 10; Toiduainetööstuse üld- ja finantsandmed

Studies on top-quark Monte Carlo modelling for Top2016

CERN Document Server

The ATLAS collaboration

2016-01-01

This note summarises recent studies on Monte Carlo simulation setups of top-quark pair production used by the ATLAS experiment and presents a new method to deal with interference effects for the $Wt$ single-top-quark production which is compared against previous techniques. The main focus for the top-quark pair production is on the improvement of the modelling of the Powheg generator interfaced to the Pythia8 and Herwig7 shower generators. The studies are done using unfolded data at centre-of-mass energies of 7, 8, and 13 TeV.

Feedback from dismantling operations (level 2) on EDF's first generation reactors

International Nuclear Information System (INIS)

West, J P.; Dionisio-Gomes, A.; Kus, J P.; Mervaux, P.; Bernet, P.; Dalmas, R.

2003-01-01

EDF's policy as regards the dismantling of the reactors that have ceased commercial operation, namely the eight power plants of the first generation and the Creys-Malville power plant, is explained. Generally speaking, prior to the year 2001, EDF had opted for the de-construction of these power plants to comply with a 'long wait' scenario, which consisted of waiting for a period of 5 to 10 years to achieve IAEA level 2 (partial release of the site), then postponing the total de-construction of the facilities for 25 to 50 years. Today, EDF has decided to undertake the total de-construction of these reactors, which have ceased commercial operation, over a period of 25 years. The purpose of this document is to present: - The reactors concerned, their background and their 'regulatory' situation, - The main operations performed and/or currently in progress, - The main elements of feedback from such operations, shedding light on the approach adopted in 2001. The installations concerned by the de-construction programme are as follows: - The 8 power plants of the first generation, which were built during the fifties and sixties and ceased commercial operation between 1973 and 1994, namely: Brennilis (industrial prototype using heavy water technology, jointly operated by EDF and CEA), the 6 power units of the NUGG type (natural uranium gas graphite) at Chinon, Saint-Laurent des Eaux and Bugey and the PWR reactor at Chooz A, - The storage silos at Saint-Laurent, where the sleeves for the fuel assemblies of reactors SLA1 and SLA2 are stored, corresponding to approximately 2000 tonnes of graphite, - The Creys-Malville reactor, FBR (fast breeder reactor) shut down in accordance with a government decision, which is currently undergoing decommissioning. At the current stage, our feedback from the dismantling operations carried out on nuclear facilities is based on (i) the work carried out or in progress that will make it possible to achieve the equivalent of IAEA level 2 in the

Removable top nozzle and tool for a nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Wilson, J.F.; Cerni, S.; Gjertsen, R.K.

1987-01-01

A tool is described used in a nuclear reactor fuel assembly for connecting and disconnecting an internally threaded collar threadably engagable with an externally threaded upper end of a control rod guide thimble. The tool consists of: (a) rotation means engagable with the collar and operable to apply a torque in one direction to threadably connect the internally threaded collar on the thimble's externally threaded upper end and in an opposite direction to disconnect the collar from its threaded connection with the thimble; and (b) gripper means adapted to be inserted into the upper end of the guide thimble and operable to prevent the thimble from rotating about its longitudinal axis as the rotation means applies torque to the collar in connecting and disconnecting the collar on and off the guide thimble; (c) the gripper means being disposed within the rotation means and having a portion thereof projecting outwardly beyond the rotation means for insertion into the thimble when the rotation means is engaged with the collar

Top-philic scalar Dark Matter with a vector-like fermionic top partner

OpenAIRE

Baek, Seungwon; Ko, Pyungwon; Wu, Peiwen

2016-01-01

We consider a simple extension of the Standard Model with a scalar top-philic Dark Matter (DM) $S$ coupling, apart from the Higgs portal, exclusively to the right-handed top quark $t_R$ and a colored vector-like top partner $T$ with a Yukawa coupling $y_{ST}$ which we call the topVL portal. When the Higgs portal is closed and $y_{ST}$ is perturbative $ (\\lesssim 1)$, $TS\\to (W^+b, gt)$, $SS\\to t\\bar{t}$ and $T\\bar{T}\\to (q\\bar{q},gg)$ provide the dominant (co)annihilation contributions to obt...

Assessment of extent and degree of thermal damage to polymeric materials in the Three Mile Island Unit 2 reactor building

International Nuclear Information System (INIS)

Alvares, N.J.

1986-01-01

This paper describes assumptions and procedures used to perform thermal damage analysis caused by post loss-of-coolant-accident (LOCA) hydrogen deflagration at Three Mile Island Unit 2 Reactor. Examination of available photographic evidence yields data on the extent and range of thermal and burn damage. Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, the authors assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. A control pendant from the polar crane located in the top of the reactor building sustained asymmetric burn damage of decreasing degree from top to bottom. Evidence suggests the polar-crane pendant side that experienced heaviest damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Simple hydrogen-fire-exposure tests and heat transfer calculations approximate the degree of damage found on inspected materials from the containment building and support for an estimated 8% pre-fire hydrogen

Assessment of extent and degree of thermal damage to polymeric materials in the Three Mile Island Unit 2 Reactor building

International Nuclear Information System (INIS)

Alvares, N.J.

1985-06-01

This paper describes assumptions and procedures used to perform thermal damage analysis caused by post loss-of-coolant-accident (LOCA) hydrogen deflagration at Three Mile Island Unit 2 Reactor. Examination of available photographic evidence yields data on the extent and range of thermal and burn damage. Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, we assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. A control pendant from the polar crane located in the top of the reactor building sustained asymmetric burn damage of decreasing degree from top to bottom. Evidence suggests the polar-crane pendant side that experienced heaviest damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Simple hydrogen-fire-exposure tests and heat transfer calculations approximate the degree of damage found on inspected materials from the containment building and support for an estimated 8% pre-fire hydrogen

Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

International Nuclear Information System (INIS)

Forsberg, C.

2015-01-01

The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

International Nuclear Information System (INIS)

Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

2007-01-01

In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

Evolution of on-power fuelling machines on Canadian natural uranium power reactors

International Nuclear Information System (INIS)

Isaac, P.

1984-10-01

The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

Ida-Virumaa ettevõtete TOP 50

Index Scriptorium Estoniae

2005-01-01

Ilmunud ka: Delovõje Vedomosti-Severo-Vostok 7. sept. lk. 4. Ida-Virumaa ettevõtete TOP 50; Käibe TOP 35; Kasumi TOP 35; Käibe kasvu TOP 20; Kasumi kasvu TOP 20; Rentaabluse TOP 20; Omakapitali tootlikkuse TOP 20; Ida-Virumaa ettevõtete üld- ja finantsandmed

Raising the four downcomers in the reactor aluminium tank of the FRJ-2 research reactor as an example of the execution of complicated work in the region of high radiation levels

International Nuclear Information System (INIS)

Nickel, M.; Schmitz, J.; Wolters, J.

1975-02-01

As a result of the planned power increase from 15 MW to 25 MW, a new emergency cooling system had to be installed in the research reactor FRJ-2 of the KFA Juelich, which called for an extension of the four standpipes in the reactor tank by 57 mm. Due to the high radiation level in the reactor tank, new techniques had to be found allowing aluminium rings of corresponding height to be welded onto the top part of the standpipes by remotecontrolled welding; moreover, the welded parts were then to be protected by a bandage made of high-quality steel. The development work was carried out in the KFA and this report gives an account of the technique applied and the results obtained. (author)

Safety of intrinsically safe and economical reactor (ISER)

International Nuclear Information System (INIS)

Asahi, Y.; Sugawara, I.; Yamanaka, K.

1988-01-01

Inherent safety of a reactor may be quantified by the grace period at various safety levels such as maintenance of fuel integrity, maintenance of fuel coolability and avoidance of core-melt. It is important to find out the grace period especially at the safety level of maintenance of fuel integrity. It has been conducted to design the ISER, which is characterized by the steel-made reactor pressure vessel. In addition to the passive nature of the safety design of the reactor itself, the ISER is equipped in the secondary system with a subsystem called the passive safety and shutdown system (PSSS), which will help to increase the grace period. It was found by the null transient analysis that check valves are needed at the top hot/cold interface. The analysis of the station blackout, which is one of the severest accident conceivable for the ISER, was made to examine inherent safety of the ISER with and without the PSSS. This paper reports that found out that the PSSS enhances inherent safety of the ISER

The search for the top quark

International Nuclear Information System (INIS)

Barbaro-Galtieri, A.

1989-03-01

This paper discusses the following topics: top search in the near future, general remarks, top search at HERA, searching for the top quarks at the Z 0 machines, finding the top at Lep II, top search in UA2, top search in UA1, and top search at CDF. 58 refs., 38 figs

Coolant clean-up system in the primary coolant circuit for nuclear reactor

International Nuclear Information System (INIS)

Saito, Michio.

1981-01-01

Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

TOP LINAC design; Progetto del TOP LINAC

Energy Technology Data Exchange (ETDEWEB)

Picardi, L; Ronsivalle, C; Vignati, A [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Innovazione

1997-11-01

The report describes a linear accelerator for protons named TOP LINAC designed for the TOP (Terapia Oncologica con Protoni, Oncological Protontherapy) project launched by the Italian National Institute of Health (Istituto Superiore di Sanita`, ISS) to explore in collaboration with the biggest Oncological Hospital in Rome (Istituto Regina Elena, IRE) the potentialities of the therapy with accelerated protons and establish guide lines for the application of this new type of radiotherapy in comparison with the more traditional electron and x-rays radiotherapy. The concept of a compact accelerator for protontherapy applications bore within the Italian Hadrontherapy Collaboration (TERA Collaboration) with the aim to diffuse the protontherapy on the National territory. The ISS program plans to use the TOP linac proton beam also for production of PET (Positron Emission Tomography) radioisotopes and radiobiology studies. Official agreements are in course between ISS and ENEA which provides its experience in the industrial and medical accelerators for the design and the construction of the TOP linac. The accelerator that will be the first 3 GHz proton linac in the world, will be composed of a 428.3 MHz 7 Me V RFQ + DTL injector followed by a 7-65 Me V section of a 3 GHz SCDTL structure and a 65 - 200 Me V variable energy SCL 3 GHz structure. In particular the SCDTL section uses a highly innovative accelerating structure patented by ENEA. In this report the clinical and physical requests are discussed and a preliminary design of the whole machine is given.

Top quark mass measurement

International Nuclear Information System (INIS)

Maki, Tuula; Helsinki Inst. of Phys.; Helsinki U. of Tech.

2008-01-01

The top quark is the heaviest elementary particle. Its mass is one of the fundamental parameters of the standard model of particle physics, and an important input to precision electroweak tests. This thesis describes three measurements of the top-quark mass in the dilepton decay channel. The dilepton events have two neutrinos in the final state; neutrinos are weakly interacting particles that cannot be detected with a multipurpose experiment. Therefore, the signal of dilepton events consists of a large amount of missing energy and momentum carried off by the neutrinos. The top-quark mass is reconstructed for each event by assuming an additional constraint from a top mass independent distribution. Template distributions are constructed from simulated samples of signal and background events, and parameterized to form continuous probability density functions. The final top-quark mass is derived using a likelihood fit to compare the reconstructed top mass distribution from data to the parameterized templates. One of the analyses uses a novel technique to add top mass information from the observed number of events by including a cross-section-constraint in the likelihood function. All measurements use data samples collected by the CDF II detector

Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

International Nuclear Information System (INIS)

Hodge, S.A.

1990-01-01

Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

Pressurized-water coolant nuclear reactor steam generator

International Nuclear Information System (INIS)

Mayer, H.; Schroder, H.J.

1975-01-01

A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

Shock loading of reactor vessel following hypothetical core disruptive accident

International Nuclear Information System (INIS)

Srinivas, G.; Doshi, J.B.

1990-01-01

Hypothetical Core Disruptive Accident (HCDA) has been historically considered as the maximum credible accident in Fast Breeder Reactor systems. Environmental consequences of such an accident depends to a great extent on the ability of the reactor vessel to maintain integrity during the shock loading following an HCDA. In the present paper, a computational model of the reactor core and the surrounding coolant with a free surface is numerical technique. The equations for conservation of mass, momentum and energy along with an equation of state are considered in two dimensional cylindrical geometry. The reactor core at the end of HCDA is taken as a bubble of hot, vaporized fuel at high temperature and pressure, formed at the center of the reactor vessel and expanding against the surrounding liquid sodium coolant. The free surface of sodium at the top of the vessel and the movement of the core bubble-liquid coolant interface are tracked by Marker and Cell (MAC) procedure. The results are obtained for the transient pressure at the vessel wall and also for the loading on the roof plug by the impact of the slug of liquid sodium. The computer code developed is validated against a benchmark experiment chosen to be ISPRA experiment reported in literature. The computer code is next applied to predict the loading on the Indian Prototype Fast Breeder Reactor (PFBR) being developed at Kalpakkam

Top-quark mass and top-quark pole mass measurements with the ATLAS detector

CERN Document Server

Barillari, Teresa; The ATLAS collaboration

2017-01-01

Results of top-quark mass measurements in the di-lepton and in the all-jets top-antitop decay channels with the ATLAS detector are presented. The measurements are obtained using proton--proton collisions at a centre-of-mass energy \\sqrt{s} = 8 TeV at the CERN Large Hadron Collider. The data set used corresponds to an integrated luminosity of 20.2 fb-1. The top-quark mass in the di-lepton channel is measured to be 172.99 +/-0.41 (stat.) +/- 0.74 (syst.) GeV. In the all-jets analysis the top-quark mass is measured to be 173.72 +/- 0.55 (stat.)+/- 1.01 (syst.) GeV. In addition, the top-quark pole mass is determined from inclusive cross-section measurements in the top-antitop di-lepton decay channel with the ATLAS detector. The measurements are obtained using data at \\sqrt{s} = 7 TeV and \\sqrt{s} =8 TeV corresponding to an integrated luminosity of 4.6 fb-1 and 20.2 fb-1 respectively. The top-quark pole mass is measured to be 172.9^{+2.5}_{-2.6} GeV.

Sludge granulation in an UASB-moving bed biofilm hybrid reactor for efficient organic matter removal and nitrogen removal in biofilm reactor.

Science.gov (United States)

Chatterjee, Pritha; Ghangrekar, M M; Rao, Surampalli

2018-02-01

A hybrid upflow anaerobic sludge blanket (UASB)-moving bed biofilm (MBB) and rope bed biofilm (RBB) reactor was designed for treatment of sewage. Possibility of enhancing granulation in an UASB reactor using moving media to improve sludge retention was explored while treating low-strength wastewater. The presence of moving media in the top portion of the UASB reactor allowed a high solid retention time even at very short hydraulic retention times and helped in maintaining selection pressure in the sludge bed to promote formation of different sized sludge granules with an average settling velocity of 67 m/h. These granules were also found to contain plenty of extracellular polymeric substance (EPS) such as 58 mg of polysaccharides (PS) per gram of volatile suspended solids (VSS) and protein (PN) content of 37 mg/g VSS. Enriched sludge of nitrogen-removing bacteria forming a porous biofilm on the media in RBB was also observed in a concentration of around 894 g/m 2 . The nitrogen removing sludge also had a high EPS content of around 22 mg PS/g VSS and 28 mg PN/g VSS. This hybrid UASB-MBB-RBB reactor with enhanced anaerobic granular sludge treating both carbonaceous and nitrogenous matter may be a sustainable solution for decentralized sewage treatment.

Top quark production at the LHC (single top and tt-bar cross sections)

International Nuclear Information System (INIS)

Lange, J.

2014-01-01

With the large number of top quarks produced at the LHC, top quark physics enters an era of precision and properties measurements. This article reviews the recent advances in top quark cross section measurements performed by ATLAS and CMS using data recorded in 2011 with integrated luminosities up to 5 fb -1 . They include precision inclusive cross sections, the establishment of challenging channels, first differential cross section measurements and single top production. An overall good agreement with Standard Model predictions is observed

Top Physics at Atlas

CERN Document Server

Romano, M; The ATLAS collaboration

2013-01-01

This talk is an overview of recent results on top-quark physics obtained by the ATLAS collaboration from the analysis of p-p collisions at 7 and 8 TeV at the Large Hadron Collider. Total and differential top pair cross section, single top cross section and mass measurements are presented.

Fission reactor control rod

International Nuclear Information System (INIS)

Irie, Tomoo.

1991-01-01

The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

Top-down versus bottom-up processing of influence diagrams in probabilistic analysis

International Nuclear Information System (INIS)

Timmerman, R.D.; Burns, T.J.; Dodds, H.L. Jr.

1986-01-01

Recent work by Phillips et al and Selby et al has shown that influence diagram methodology can be a useful analytical tool in reactor safety studies. In some instances, an influence diagram can be used as a graphical representation of probabilistic dependence within a system or event sequence. Under these circumstances, Bayesian statistics is employed to transform the relationships depicted in the influence diagram into the correct expression for a desired marginal probability (e.g., the top node). In the references cited above, the authors demonstrated the usefulness of influence diagrams for assessing the reliability of operator performance during pressurized thermal shock transients. In addition, the use of influence diagrams identified the critical variables that had the greatest impact on operator reliability for a particular scenario (e.g., control room design, procedures, etc.). Top-down and bottom-up algorithms have emerged as the dominant methods for quantifying influence diagrams. The purpose of this paper is to demonstrate a potential error in employing the bottom-up algorithm when dealing with interdependencies

The efficiency of two anaerobic reactor components; Eficiencias de dos componentes de un reactor anaerobio

Energy Technology Data Exchange (ETDEWEB)

Vazquez Borges, E.; Mendez Novelo, R.; Magana Pietra, A. [Facultad de Ingenieria. Universidad de Yucatan (Mexico); Martinez Pereda, P.; Fernandez Villagomez, G. [Universidad Nacional Autonoma de Mexico. Division de Estudios de posgrado de la Facultad de Ingenieria. Mexico (Mexico)

1997-09-01

This study examined the behaviour of an anaerobic digester in treating pig farm sewage. The experimental model consisted of a UASB reactor at the bottom and a high-rate sedimentator at the top with a total capacity of 534 litres. The digester was installed on a pig farm and its performance under different operating conditions was determined, with hydraulic retention time (HRT) as the critical parameter for evaluating the anaerobic system`s efficiency. The results obtained during the experiment to establish the critical operating parameters are reported. The organic loads applied for a HRT of 1 day were 7.3 kg/m``3/day of total DQO and 3 kg/m``3/day of soluble DQO, following organic matter removal rates (as total DQO) of 36% and 49% respectively and removal rates (as soluble DQO) of 74% in the UASB and 8% in the sedimentator. The efficiency of the reactor as a whole at this HRT time was a removal rate of 74% of total DQO and 75% of soluble DQO. (Author) 25 refs.

Progress in the development of tooling and dismantling methodologies for the Windscale advanced gas cooled reactor (WAGR)

International Nuclear Information System (INIS)

Cross, M.T.; Wareing, M.I.; Dixon, C.

1998-01-01

Decommissioning of the Windscale Advanced Gas-Cooled Reactor (WAGR) is a major UK reactor decommissioning project co-funded by the UK Government, the European Commission and Magnox Electric. WAGR was a CO 2 cooled, graphite moderated reactor which served as a test bed for the development of Advanced Gas-Cooled Reactor technology in the UK. It operated from 1963 until shutdown in 1981. AEA Technology plc are currently the Managing Agents on behalf of UKAEA for the WAGR decommissioning project and are responsible for the co-ordination of the project up to the point when the contents of the reactor core and associated radioactive materials are removed and either disposed of or packaged for disposal at some time in the future. Decommissioning has progressed to the point where the reactor has been dismantled down to the level of the hot gas collection manifold with the removal of the top biological shield, the refuelling standpipes and the top section of the reactor pressure vessel. The 4 heat exchangers have also been removed and committed to shallow land burial. This paper describes the work carried out by AEA Technology under separate contracts of UKAEA in developing some of the equipment and deployment methods for the next phase of active operations required in preparation for the dismantling of the core structure. Most recent work has concentrated on the development of specialist tooling for removal of items of operational waste stored within the reactor core, equipment for cutting and removal of the highly radioactive stainless steel 'loop' pressure tubes, diamond wire cutting equipment for sectioning large diameter pipework, and equipment for dismantling the reactor neutron shield. The paper emphasises the process of adaptation and extension of existing technologies for cost-effective application in the decommissioning environment, the need for adequate forward planning of decommissioning methodologies together with large-scale 'mock-up' testing of equipment to

Top physics at CDF

Energy Technology Data Exchange (ETDEWEB)

Hughes, R.E. [Univ. of Rochester, NY (United States)

1997-01-01

We report on top physics results using a 100 pb{sup -1} data sample of p{bar p} collisions at {radical}s = 1.8 TeV collected with the Collider Detector at Fermilab (CDF). We have identified top signals in a variety of decay channels, and used these channels to extract a measurement of the top mass and production cross section. A subset of the data (67 pb{sup -1}) is used to determine M{sub top} = 176 {+-} 8(stat) {+-} 10(syst) and {sigma}(tt) = 7.6 {sub -2.0}{sup +2.4} pb. We present studies of the kinematics of t{bar t} events and extract the first direct measurement of V{sub tb}. Finally, we indicate prospects for future study of top physics at the Tevatron.

The neutrino in all its states - Seminar dedicated to Jacques Bouchez - Slides of the presentations

International Nuclear Information System (INIS)

Spiro, M.; Pessard, H.; Rubbia, A.; Petcov, S.; Cousins, B.; Fechner, M.; Mezetto, M.

2011-01-01

The present scientific seminar, organized in the memory of Jacques Bouchez is centered on neutrino physics and presents the state of the art on experiments, on future projects and on the theory of neutrinos (oscillations and MSW effect). This document is made up of the slides of 7 presentations: 1) The achievements of J.Bouchez; 2) Reactor neutrino experiments from Bugey to double-Chooz (via RENO and Daya-Bay); 3) Neutrinos and accelerators: on the way toward the third flavor (NOMA, OPERA and T2K experiments); 4) Neutrino oscillations and MSW effect; 5) Some statistical questions in neutrino physics; 6) Long baseline oscillations: towards Japan future neutrino oscillation experiments; and 7) Next generation of neutrino oscillation experiments. (A.C.)

Large short-baseline νμ disappearance

International Nuclear Information System (INIS)

Giunti, Carlo; Laveder, Marco

2011-01-01

We analyze the LSND, KARMEN, and MiniBooNE data on short-baseline ν μ →ν e oscillations and the data on short-baseline ν e disappearance obtained in the Bugey-3 and CHOOZ reactor experiments in the framework of 3+1 antineutrino mixing, taking into account the MINOS observation of long-baseline ν μ disappearance and the KamLAND observation of very-long-baseline ν e disappearance. We show that the fit of the data implies that the short-baseline disappearance of ν μ is relatively large. We obtain a prediction of an effective amplitude sin 2 2θ μμ > or approx. 0.1 for short-baseline ν μ disappearance generated by 0.2 2 2 , which could be measured in future experiments.

Monitoring of fission products through on-line gamma spectrometry

International Nuclear Information System (INIS)

Montagnon, F.; Warlop, R.

1989-01-01

Under normal operating conditions, the monitoring of the possible deterioration of the pressurized water reactor core fuel rods is achieved through analysis of the radioactive fission products carried by the primary system. For acquiring results of spectrometric analyses in real time, and avoiding risks of errors linked to manual operations, CEA/DMG and EDF/SEPTEN have jointly developed an entirely automatic system. This system allows measuring permanently the primary system activity of two coupled units, with no human operation nor any handling of active coolant specimens. The PIGAL facility has been set up in the nuclear auxiliary building, common to the two units, and it is used on a demonstration basis for units 2 and 3 of the BUGEY site. This device has been patented

Entangling Higgs production associated with a single top and a top-quark pair in the presence of anomalous top-Yukawa coupling

Energy Technology Data Exchange (ETDEWEB)

Chang, Jung [Physics Division, National Center for Theoretical Sciences,Hsinchu, Taiwan (China); Cheung, Kingman [Physics Division, National Center for Theoretical Sciences,Hsinchu, Taiwan (China); Division of Quantum Phases and Devices, School of Physics, Konkuk University,Seoul 143-701 (Korea, Republic of); Department of Physics, National Tsing Hua University,Hsinchu 300, Taiwan (China); Lee, Jae Sik [Physics Division, National Center for Theoretical Sciences,Hsinchu, Taiwan (China); Department of Physics, Chonnam National University, 300 Yongbong-dong, Buk-gu, Gwangju, 500-757 (Korea, Republic of); Lu, Chih-Ting [Department of Physics, National Tsing Hua University,Hsinchu 300, Taiwan (China)

2017-04-26

The ATLAS and CMS collaborations observed a mild excess in the associated Higgs production with a top-quark pair (tt̄h) and reported the signal strengths of μ{sub tth}{sup ATLAS}=1.81±0.80 and μ{sub tth}{sup CMS}=2.75±0.99 based on the data collected at √s= 7 and 8 TeV. Although, at the current stage, there is no obvious indication whether the excess is real or due to statistical fluctuations, here we perform a case study of this mild excess by exploiting the strong entanglement between the associated Higgs production with a single top quark (thX) and tt̄h production in the presence of anomalous top-Yukawa coupling. As well known, tt̄h production only depends on the absolute value of the top-Yukawa coupling. Meanwhile, in thX production, this degeneracy is lifted through the strong interference between the two main contributions which are proportional to the top-Yukawa and the gauge-Higgs couplings, respectively. Especially, when the relative sign of the top-Yukawa coupling with respect to the gauge-Higgs coupling is reversed, the thX cross section can be enhanced by more than one order of magnitude. We perform a detailed study of the influence of thX production on tt̄h production in the presence of the anomalous top-Yukawa coupling and point out that it is crucial to include thX production in the analysis of the tt̄h data to pin down the sign and the size of the top-Yukawa coupling in future. While assuming the Standard Model (SM) value for the gauge-Higgs coupling, we vary the top-Yukawa coupling within the range allowed by the current LHC Higgs data. We consider the Higgs decay modes into multileptons, bb̄ and γγ putting a particular emphasis on the same sign dilepton events. We also discuss the prospects for the LHC Run-2 on how to disentangle thX production from tt̄h one and how to probe the anomalous top-Yukawa coupling.

The AFEN Method in Cylindrical (r,θ ,z) Geometry for Pebble Bed Reactors -Incorporation of Acceleration and Discontinuity Factor

International Nuclear Information System (INIS)

Lee, Jaejun; Cho, Namzin

2007-01-01

Most existing methods of nuclear design analysis for pebble bed reactors (PBRs) are based on old finite difference solvers or on statistical methods. These methods require very long computer times. Therefore, there is strong desire of making available high fidelity coarse-mesh nodal computer codes. Recently, we extended the analytic function expansion nodal (AFEN) method developed quite extensively in Cartesian (x,y,z) geometry and in hexagonal-z geometry to the treatment of the full three dimensional cylindrical (r,θ,z) geometry for pebble bed reactors(PBRs). The AFEN methodology in this geometry as in hexagonal geometry is 'robust', due to the unique feature of the AFEN method that it does not use the transverse integration. This paper presents an acceleration scheme based on the coarse-group rebalance (CGR) concept and provides test results verifying the method and its implementation in the TOPS code. Also, we implemented discontinuity factors in the TOPS code and tested on benchmark problems. The TOPS results are in excellent agreement with those of the VENTURE code, using significantly less computer time

Anomalous couplings in single top and searches for rare top quark couplings with the ATLAS detector

CERN Document Server

Cabrera Urban, Susana; The ATLAS collaboration

2017-01-01

The top quark is the heaviest known fundamental particle and probing its couplings with the other fundamental particle may open a window to physics beyond the Standard Model. Single top-quark production provides a unique window to study the coupling between the top quark, the W boson and the b quark, since it involves the $Wtb$ vertex in both production and decay. Measurements of angular correlations in single top quark events in the t-channel exchange of a W boson are presented based on the 8 TeV ATLAS dataset. Differential cross-sections are measured as a function of angular variables that are sensitive to anomalous contributions to the Wtb vertex and the top quark polarization. Searches for flavour-changing neutral current top-quark interactions are also discussed based on the 8 TeV and 13 TeV ATLAS dataset. Searches for rare top quark decays to Higgs and Z bosons are presented in top quark production, and searches for rare top quark interactions with gluons and Z bosons are presented in single top quark p...

Wave Engine Topping Cycle Assessment

Science.gov (United States)

Welch, Gerard E.

1996-01-01

The performance benefits derived by topping a gas turbine engine with a wave engine are assessed. The wave engine is a wave rotor that produces shaft power by exploiting gas dynamic energy exchange and flow turning. The wave engine is added to the baseline turboshaft engine while keeping high-pressure-turbine inlet conditions, compressor pressure ratio, engine mass flow rate, and cooling flow fractions fixed. Related work has focused on topping with pressure-exchangers (i.e., wave rotors that provide pressure gain with zero net shaft power output); however, more energy can be added to a wave-engine-topped cycle leading to greater engine specific-power-enhancement The energy addition occurs at a lower pressure in the wave-engine-topped cycle; thus the specific-fuel-consumption-enhancement effected by ideal wave engine topping is slightly lower than that effected by ideal pressure-exchanger topping. At a component level, however, flow turning affords the wave engine a degree-of-freedom relative to the pressure-exchanger that enables a more efficient match with the baseline engine. In some cases, therefore, the SFC-enhancement by wave engine topping is greater than that by pressure-exchanger topping. An ideal wave-rotor-characteristic is used to identify key wave engine design parameters and to contrast the wave engine and pressure-exchanger topping approaches. An aerodynamic design procedure is described in which wave engine design-point performance levels are computed using a one-dimensional wave rotor model. Wave engines using various wave cycles are considered including two-port cycles with on-rotor combustion (valved-combustors) and reverse-flow and through-flow four-port cycles with heat addition in conventional burners. A through-flow wave cycle design with symmetric blading is used to assess engine performance benefits. The wave-engine-topped turboshaft engine produces 16% more power than does a pressure-exchanger-topped engine under the specified topping

Self-pressurization analysis of the natural circulation integral nuclear reactor using a new dynamic model

Directory of Open Access Journals (Sweden)

Ali Farsoon Pilehvar

2018-06-01

Full Text Available Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established. Keywords: Condensation Power, Flashing Phenomenon, Natural Circulation, Self-Pressurization, Small Modular Reactor

CPT analysis with top physics

Energy Technology Data Exchange (ETDEWEB)

Cembranos, Jose A. R., E-mail: cembra@fis.ucm.es [Universidad Complutense de Madrid, Departamento de Fisica Teorica I (Spain)

2013-03-15

We discuss the possibility of observing CPT violation from top anti-top production in hadronic colliders. We study a general approach by analyzing constraints on the mass difference between the top and anti-top quarks. We present current bounds from Tevatron data, and comment on the prospects for improving these bounds at the LHC and the ILC.

Top Quark Properties at Tevatron

Energy Technology Data Exchange (ETDEWEB)

Lysák, Roman [Prague, Inst. Phys.

2017-11-27

The latest CDF and D0 experiment measurements of the top quark properties except the top quark mass are presented. The final combination of the CDF and D0 forward-backward asymmetry measurements is shown together with the D0 measurements of the inclusive top quark pair cross-section as well as the top quark polarization.

System for bearing a nuclear reactor vessel cooled by liquid metal

International Nuclear Information System (INIS)

Mahe, A.; Jullien, G.

1976-01-01

The invention relates to a bearing system for supporting a nuclear reactor vessel of the kind which is suspended from the reactor closure slab. The bearing system comprises a ring connected at one end to a collar and at the other end to two collars. The collar connected to the bottom end of the ring forms the top part of the vessel to be supported while the other two collars fit into the slab at two separate places. The ring and collars are disposed in an annular space formed in the slab and dividing it into two parts, i.e., a central part and a peripheral part surrounding the central part of the slab

Ida-Virumaa ettevõtete TOP aastal 2002

Index Scriptorium Estoniae

2003-01-01

Ettevõtete TOP 55. TOP-i esikümme. Käibe TOP 40. Kasumi TOP 40. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Ida-Virumaa firmade üldandmed. Ida-Virumaa firmade finantsandmed

Single top t-channel

CERN Document Server

Faltermann, Nils

2017-01-01

The production of single top quarks allows to study the interplay of top quark physics and the electroweak sector of the standard model. Deviations from predictions can be a hint for physics beyond the standard model. The t-channel is the dominant production mode for single top quarks at the LHC. This talk presents the latest measurements from the ATLAS and CMS collaborations.

Managerial improvement efforts after finding unreported cracks in reactor components

International Nuclear Information System (INIS)

Kawamura, S.

2006-01-01

In 2002 TEPCO found that there were unreported cracks in reactor components, of which inspection records had been falsified. Stress Corrosion Cracking indications found in Core Shrouds and Primary Loop Re-circulation pipes at some plants were removed from the inspection records and not reported to the regulators. Top management of TEPCO took the responsibility and resigned, and recovery was started under the leadership of new management team. First of all, behavioral standards were reconstituted to strongly support safety-first value. Ethics education was introduced and corporate ethics committee was organized with participation of external experts. Independent assessment organization was established to enhance quality assurance. Information became more transparent through Non-conformance Control Program. As for the material management, prevention and mitigation programs for the Stress Corrosion Cracking of reactor components were re-established. In addition to the above immediate recovery actions, long term improvement initiatives have also been launched and driven by our aspiration to excellence in safe operation of nuclear power plants. Vision and core values were set to align the people. Organizational learning was enhanced by benchmark studies, better systematic use of operational experience, self-assessment and external assessment. Based on these foundation blocks and with strong sponsorship from the top management, work processes were analyzed and improved by Peer Groups. (author)

Implementation of ALARA at the design stage of Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Brissaud, A.; Ridoux, P. [Electricite de France, Villeurbanne (France)

1995-03-01

In the 1970s, Electricite de France (EdF) had limited knowledge and experience of pressurized water reactors (PWRs). Electricity generation by nuclear units was oriented towards gas-graphite reactors, even though EdF had a share in the PWR unit of CHOOZ A-1 (250 MWe, later upgraded to 320 MWe). Some facts about the origin of doses in that king of reactor were known to the research and development (R&D) support staff of EdF, which mainly comprises the French Atomic Commission (CEA), but only a few of EdF`s engineers were aware of these facts. One has to bear in mind that CHOOZ A-1 only went critical in April 1967 and was officially connected to the grid in May 1970 after some important problems had been solved. Meanwhile, the nuclear program was launched at full speed, beginning with the order for FESSENHEIM 1 in 1970, FESSENHEIM 2 and BUGEY 2 and 3 in 1971. TIHANGE 1, in which EdF had a share, went on-line in September 1975. Also, supposing that EdF had had such knowledge and experience, it is quite evident that it would have been very difficult to modify the lay-out inside the reactor building.

Determination of the top-quark mass from hadro-production of single top-quarks

International Nuclear Information System (INIS)

Alekhin, S.; Moch, S.; Thier, S.

2016-08-01

We present a new determination of the top-quark mass m_t based on the experimental data from the Tevatron and the LHC for single-top hadro-production. We use the inclusive cross sections of s- and t-channel top-quark production to extract m_t and to minimize the dependence on the strong coupling constant and the gluon distribution in the proton compared to the hadro-production of top-quark pairs. As part of our analysis we compute the next-to-next-to-leading order approximation for the s-channel cross section in perturbative QCD based on the known soft-gluon corrections and implement it in the program HatHor for the numerical evaluation of the hadronic cross section. Results for the top-quark mass are reported in the MS and in the on-shell renormalization scheme.

Refueling the RPI reactor facility with low-enrichment fuel

International Nuclear Information System (INIS)

Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

1985-01-01

The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

Top quark theory

NARCIS (Netherlands)

Laenen, E.

2012-01-01

The theoretical aspects of a number of top quark properties such as its mass and its couplings are reviewed. Essential aspects in the theoretical description of top quark production, singly, in pairs and in association, as well as its decay related to spin and angular correlations are discussed.

The race to decipher the top secrets of TOP mRNAs.

Science.gov (United States)

Meyuhas, Oded; Kahan, Tamar

2015-07-01

Cells encountering hostile growth conditions, like those residing in the middle of a newly developing solid tumor, conserve resources and energy by downregulating protein synthesis. One mechanism in this response is the translational repression of multiple mRNAs that encode components of the translational apparatus. This coordinated translational control is carried through a common cis-regulatory element, the 5' Terminal OligoPyrimidine motif (5'TOP), after which these mRNAs are referred to as TOP mRNAs. Subsequent to the initial structural and functional characterization of members of this family, the research of TOP mRNAs has progressed in three major directions: a) delineating the landscape of the family; b) establishing the pathways that transduce stress cues into selective translational repression; and c) attempting to decipher the most proximal trans-acting factor(s) and defining its mode of action--a repressor or activator. The present chapter critically reviews the development in these three avenues of research with a special emphasis on the two "top secrets" of the TOP mRNA family: the scope of its members and the identity of the proximal cellular regulator(s). This article is part of a Special Issue entitled: Translation and Cancer. Copyright © 2014 Elsevier B.V. All rights reserved.

Võrumaa ettevõtete TOP 50 aastal 2003

Index Scriptorium Estoniae

2004-01-01

Võrumaa ettevõtete TOP 50 aastal 2003. Käibe TOP 40. Kasumi TOP 40. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Võrumaa firmade üld- ja finantsandmed

Top quark production at the LHC

CERN Document Server

Gilles, Geoffrey; The ATLAS collaboration

2018-01-01

The top quark is the heaviest known fundamental particle. As it is the only quark that decays before it hadronizes, it gives us the unique opportunity to probe the properties of bare quarks at the Large Hadron Collider. This talk will present highlights of a few recent precision measurements of the top quark using 13 TeV and 8 TeV collision data: top-quark pair and single top production cross sections, including differential distributions and production in association with bosons, will be presented alongside top quark properties measurements. These measurements, including results using boosted top quarks, probe our understanding of top quark production in the TeV regime. Measurements of the top quark mass are also presented.

Chernobyl: disinformation. 2000 dead in Chernobyl: they were made by journalists

International Nuclear Information System (INIS)

1996-01-01

After having briefly recalled the scenario which resulted in the Chernobyl accident, the differences between the Chernobyl reactor and the French graphite-based reactors, and that some modifications have been introduced in the alarm system of the Bugey reactor, the author reviews the chronology of events and information after the accident. He recalls the Soviet way to deal with the information, states that the authorities did not immediately understand the severity of the accident. He notably outlines and comments a statement reporting that 2.000 people died just after the accident. Thus, it appears that both sides could be criticized, the Russian side for its slow reaction, and the Western side for disinformation. He also denounces a fake documentary report made by a French journalist (images pretended to have been taken in Chernobyl had been in fact shot in an Italian factory). He also evokes the reactions of people in front of this kind information about risks of exposure. He analyses the content of an article written by a French journalist who denounced some kinds of plots elaborated by oil companies, by the USSR, by anti-nuclear activists

Läänemaa ettevõtete TOP 50

Index Scriptorium Estoniae

2005-01-01

Läänemaa ettevõtete TOP 50; Käibe TOP 35; Kasumi TOP 35; Käibe kasvu TOP 20; Kasumi kasvu TOP 20; Rentaabluse TOP 20; Omakapitali tootlikkuse TOP 20; Läänemaa ettevõtete üld- ja finantsandmed

Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

Energy Technology Data Exchange (ETDEWEB)

Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

Hydraulic stud-tensioning machines in reactor technology

International Nuclear Information System (INIS)

Lachner, H.

1978-01-01

Hydraulic multiple stud tensioner (MST) for the simultaneous prestressing of all the stud bolts is make it possible to achieve highly accurate prestress levels in the highly stressed bolts holding down the top head of reactor pressure vessels. These machines can remove and replace the nuts and studs, and can rotate these components upwards and downwards, during the operation of opening and closing the reactor pressure vessel. In order to reduce the radiation exposure of the service personnel, and also to reduce the time required for this work which may lie in the critical path of the refuelling time schedule, it is desirable to achieve complete mechanisation of these machines, including remote control and remote monitoring. The devices and components required for this purpose are without precedent in machine construction with respect to their functions and to the load range involved. The reported operating experience therefore also covers some points of general interest while the data on maintenance reflect the known status of the technology. (orig.) [de

A Preliminary Neutral Framework for the Accident Sequence Evaluation for a Hydrogen Conversion Reactor

International Nuclear Information System (INIS)

Han, Seok Jung; Yang, Joon Eon

2005-01-01

A framework for an early stage PSA for a hydrogen conversion reactor has been proposed in this paper. The approach is based on a functional and top-down approach. A main concerning point of this approach is to use a design neutral framework. A design neutral framework of PSA can provide a flexibility to apply to several candidate design concepts or options. This neutral-framework idea was borrowed from a proposed regulatory framework in US NRC. The feasibility of our proposed approach has been assessed to be applied in an accident sequence analysis for a hydrogen conversion reactor

Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

International Nuclear Information System (INIS)

Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

1987-01-01

Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)

The operating reliability of the reactor coolant pump

International Nuclear Information System (INIS)

Grancy, W.

1996-01-01

There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

Experimental study of hot water layer in a model in scale of the Brazilian Multipurpose Reactor (RMB); Estudo experimental da camada de Água quente em um modelo em escala do Reator Multipropósito Brasileiro (RMB)

Energy Technology Data Exchange (ETDEWEB)

Tomaz, Gabriel Caio Queiroz

2017-07-01

The Brazilian Multipurpose Reactor (RMB) is a 30 MW open pool research reactor planned to be constructed in Brazil. Such type of reactor is built inside a deep pool of purified and demineralized water, providing radiological protection still keeping the core accessible for maintenance and refueling. However, dissolved ions become activated in the pool water due to the core neutron flux, releasing radiation in the reactor room when the activated elements reach the top. Thus high power open pool reactors, as RMB, have an auxiliary thermal-hydraulic circuit that creates a Hot Water Layer (HWL) on the pool’s top, keeping the activated water under the HWL and mitigating the dose rate to which the operators are exposed to. The Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) built a 1/10 scale experimental bench of the RMB’s pool for the HWL investigation. This work presents the results of the pool’s heating due to the reactor startup in the HWL stability. (author)

Top quark production cross-section measurements

CERN Document Server

Chen, Ye; The ATLAS collaboration

2017-01-01

Measurements of the inclusive and differential cross-sections for top-quark pair and single top production cross sections in proton-proton collisions with the ATLAS detector at the Large Hadron Collider are presented at center-of-mass energies of 8 TeV and 13 TeV. The inclusive measurements reach high precision and are compared to the best available theoretical calculations. These measurements, including results using boosted tops, probe our understanding of top-pair production in the TeV regime. The results are compared to Monte Carlo generators implementing LO and NLO matrix elements matched with parton showers and NLO QCD calculations. For the t-channel single top measurement, the single top-quark and anti-top-quark total production cross-sections, their ratio, as well as differential cross sections are also presented. A measurement of the production cross-section of a single top quark in association with a W boson, the second largest single-top production mode, is also presented. Finally, measurements of ...

Analysis of a total loss of pool water accident in MTR-type research reactors

International Nuclear Information System (INIS)

Yilmazer, A.; Yavuz, H.

2004-01-01

In this study, the transient in which the pool water is lost throughout one or more of the main coolant pipes which are supposed to be broken guillotine-like is investigated for the TR-2 research reactor in Istanbul. The applicability of the methods used for other similar types of research reactors is shown. Decrease of the pool water level until the top of the core, and from the top to the bottom of the core are examined as two successive phases of the accident. Finite difference scheme and integral methods are employed to solve energy equations and the results of both methods are compared. The finite difference solution uses an explicit form for the analysis of the first phase, and a moving boundary approach for the second phase. The integral method is based on the assumption that the temperatures appearing in the energy equations have the same profiles during the transient as the steady state ones. Analyses are done both for nominal and hot channel, and the results of both methods are observed to be in agreement. (orig.)

Analysis of a total loss of pool water accident in MTR-type research reactors

Energy Technology Data Exchange (ETDEWEB)

Yilmazer, A. [Hacettepe University, Ankara (Turkey). Nuclear Engineering Department; Yavuz, H. [Istanbul Technical University (Turkey). Energy Institute

2004-08-01

In this study, the transient in which the pool water is lost throughout one or more of the main coolant pipes which are supposed to be broken guillotine-like is investigated for the TR-2 research reactor in Istanbul. The applicability of the methods used for other similar types of research reactors is shown. Decrease of the pool water level until the top of the core, and from the top to the bottom of the core are examined as two successive phases of the accident. Finite difference scheme and integral methods are employed to solve energy equations and the results of both methods are compared. The finite difference solution uses an explicit form for the analysis of the first phase, and a moving boundary approach for the second phase. The integral method is based on the assumption that the temperatures appearing in the energy equations have the same profiles during the transient as the steady state ones. Analyses are done both for nominal and hot channel, and the results of both methods are observed to be in agreement. (orig.)

Consideration on risk reduction of future breeder reactors

International Nuclear Information System (INIS)

Vossebrecker, H.

1990-09-01

An overall concept of risk minimization of future sodium-cooled fast breeder reactors is presented in this report. Since shutdown reliability is of vital importance for the breeder safety, a so-called third shutdown level is proposed in addition to the two independent fast shutdown systems. It is basically a group of passive and active measures, which are capable to bring the reactor to safe conditions in all conceivable accident-initiating events and in case of total failure of the two actual shutdown systems. Core disruptions as a result of shutdown failure are therefore beyond the scope of technical imagination. Measures are also foreseen to combat other conceivable causes of core disruption, in particular to achieve residual heat removal with essentially passive systems by making use of the good natural circulation capacity of sodium. On top of that, since absolute safety can never be claimed, damage-limiting containment measures are discussed

Safety and licensing of MHTGR [Modular High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Silady, F.A.; Millunzi, A.C.; Kelley, A.P. Jr.; Cunliffe, J.

1987-07-01

The Modular High Temperature Gas Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant boundary. The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

Science.gov (United States)

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Monte Carlo simulation of core physics parameters of the Syrian MNSR reactor

International Nuclear Information System (INIS)

Khattab, K.; Sulieman, I.

2011-01-01

A 3-D neutronic model for the Syrian Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis using the MCNP-4C code. The continuous energy neutron cross sections were evaluated from the ENDF/B-VI library. This model is used in this paper to calculate the following reactor core physics parameters: the clean cold core excess reactivity, calibration of the control rod and calculation its shut down margin, calibration of the top beryllium shim plate reflector, the axial neutron flux distributions in the inner and outer irradiation positions and calculations of the prompt neutron life time (ι p ) and the effective delayed neutron fraction ( β e ff). Good agreements are noticed between the calculated and the measured results. These agreements indicate that the established model is an accurate representation of Syrian MNSR core and will be used for other calculations in the future. (author)

The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474

Energy Technology Data Exchange (ETDEWEB)

Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

2012-07-01

The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the

Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. 2. Replacement of upper core structure

International Nuclear Information System (INIS)

Ushiki, Hiroshi; Ito, Hiromichi; Okuda, Eiji; Suzuki, Nobuhiro; Sasaki, Jun; Oota, Katsu; Kawahara, Hirotaka; Takamatsu, Misao; Nagai, Akinori; Okawa, Toshikatsu

2015-01-01

In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of MARICO-2 (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS) in 2007. As a part of the restoration work, UCS replacement was begun at March 24, 2014 and was completed at December 17. In-vessel repair (including observation) for sodium-cooled fast reactors (SFRs) is distinct from that for light water reactors and necessitates independent development. Application of developed in-vessel repair techniques to operation and maintenance of SFRs enhanced their safety and integrity. There is little UCS replacement experience in the world and this experience and insights, which were accumulated in the replacement work of in-vessel large structure (UCS) used for more than 30 years, are expected to improve the in-vessel repair techniques in SFRs. (author)

Study on the seismic response of reactor vessel of pool type LMFBR including fluid-structure interaction

International Nuclear Information System (INIS)

Tanimoto, K.; Ito, T.; Fujita, K.; Kurihara, C.; Sawada, Y.; Sakurai, A.

1988-01-01

The paper presents the seismic response of reactor vessel of pool type LMFBR with fluid-structure interaction. The reactor vessel has bottom support arrangement, the same core support system as Super-Phenix in France. Due to the bottom support arrangement, the level of core support is lower than that of the side support arrangement. So, in this reactor vessel, the displacement of the core top tends to increase because of the core's rocking. In this study, we investigated the vibration and seismic response characteristics of the reactor vessel. Therefore, the seismic experiments were carried out using one-eighth scale model and the seismic response including FSI and sloshing were investigated. From this study, the effect of liquid on the vibration characteristics and the seismic response characteristics of reactor vessel were clarified and sloshing characteristics were also clarified. It was confirmed that FEM analysis with FSI can reproduce the seismic behavior of the reactor vessel and is applicable to seismic design of the pool type LMFBR with bottom support arrangement. (author). 5 refs, 14 figs, 2 tabs

Water chemistry management of research reactor in JAERI

Energy Technology Data Exchange (ETDEWEB)

Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

Measurement of top quark polarisation in t-channel single top quark production

OpenAIRE

CMS Collaboration

2015-01-01

Journal of High Energy Physics 2016.4 (2016): 073 reproduced by permission of Scuola Internazionale Superiore di Studi Avanzati (SISSA) Artículo escrito por un elevado número de autores, sólo se referencian el que aparece en primer lugar, el nombre del grupo de colaboración, si le hubiera, y los autores pertenecientes a la UAM A first measurement of the top quark spin asymmetry, sensitive to the top quark polarisation, in t-channel single top quark production is presented. It is based o...

Primary system thermal hydraulics of future Indian fast reactors

Energy Technology Data Exchange (ETDEWEB)

Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

2015-12-01

Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

Science.gov (United States)

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

Top quark studies at hadron colliders

Energy Technology Data Exchange (ETDEWEB)

Sinervo, P.K. [Univ. of Toronto, Ontario (Canada)

1997-01-01

The techniques used to study top quarks at hadron colliders are presented. The analyses that discovered the top quark are described, with emphasis on the techniques used to tag b quark jets in candidate events. The most recent measurements of top quark properties by the CDF and DO Collaborations are reviewed, including the top quark cross section, mass, branching fractions, and production properties. Future top quark studies at hadron colliders are discussed, and predictions for event yields and uncertainties in the measurements of top quark properties are presented.

Top quark studies at hadron colliders

International Nuclear Information System (INIS)

Sinervo, P.K.

1997-01-01

The techniques used to study top quarks at hadron colliders are presented. The analyses that discovered the top quark are described, with emphasis on the techniques used to tag b quark jets in candidate events. The most recent measurements of top quark properties by the CDF and DO Collaborations are reviewed, including the top quark cross section, mass, branching fractions, and production properties. Future top quark studies at hadron colliders are discussed, and predictions for event yields and uncertainties in the measurements of top quark properties are presented

Top quark studies at hadron colliders

International Nuclear Information System (INIS)

Sinervo, P.K.

1996-08-01

The techniques used to study top quarks at hadron colliders are presented. The analyses that discovered the top quark are described, with emphasis on the techniques used to tag b quark jets in candidate events. The most recent measurements of top quark properties by the CDF and D null collaborations are reviewed, including the top quark cross section, mass, branching fractions and production properties. Future top quark studies at hadron colliders are discussed, and predictions for event yields and uncertainties in the measurements of top quark properties are presented

Nuclear reactor plant with a gas-cooled nuclear reactor situated in a cylindrical prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Becker, G.; Elter, C.; Fritz, R.; Rautenberg, J.; Schoening, J.; Stracke, W.

1986-01-01

A simplified construction of the nuclear reactor plant with a guarantee of great safety is achieved by the auxiliary heat exhangers, which remove the post-shutdown heat in fault situations, being arranged in the wellknown way in pairs above one another in a vertical shaft. The associated auxiliary blowers are situated at the top for the upper auxiliary heat exchangers and at the bottom for the lower auxiliary heat exchangers. The cold gas is taken from the lower auxiliary blowers through a parallel gas pipe laid in concrete, which enters the vertical shaft concerned in the area of the cold gas pipe. (orig./HP) [de

Differential Top Cross-section Measurements

CERN Document Server

Fenton, Michael James; The ATLAS collaboration

2017-01-01

The top quark is the heaviest known fundamental particle. The measurement of the differential top-quark pair production cross-section provides a stringent test of advanced perturbative QCD calculations. The ATLAS collaboration has performed detailed measurements of those differential cross sections at a centre-of-mass energy of 13 TeV. This talk focuses on differential cross-section measurements in the lepton+jets final state, including using boosted top quarks to probe our understanding of top quark production in the TeV regime.

Top quark production at the Tevatron

Energy Technology Data Exchange (ETDEWEB)

Varnes, Erich W.; /Arizona U.

2010-09-01

The Fermilab Tevatron has, until recently, been the only accelerator with sufficient energy to produce top quarks. The CDF and D0 experiments have collected large samples of top quarks. We report on recent top quark production measurements of the single top and t{bar t} production cross sections, as well as studies of the t{bar t} invariant mass distribution and a search for highly boosted top quarks.

Development of a pressurizer level compensator for use on N Reactor

International Nuclear Information System (INIS)

Bussell, J.H.

1985-07-01

The instrument described in this report has been developed to compensate the measured water level in the N Reactor pressurizer for temperature effects. N Reactor is a pressurized water nuclear reactor (PWR). The instrument is defined as a pressurizer level compensator (PLC). A pressurizer is used in a PWR to control the primary coolant pressure and provide a surge volume for primary coolant expansion and contraction. A means of compensating for water and steam density is required because of the wide range of pressure and temperature that result from different steady state and transient reactor power levels. The uncompensated level is determined by measurement of differential pressure between the top of the level measurement zone and the bottom of the level measurement zone. Temperature of the water in the pressurizer is the parameter that is used to determine the proper level compensation since water and steam density are primarily functions of temperature in this case. The PLC uses a microprocessor to calculate the compensated level from temperature and differential pressure measurements. This report includes a description of the design, development, and implementation of software and hardware that are in the PLC. 9 refs., 51 figs., 17 tabs

Top Quark Physics

International Nuclear Information System (INIS)

Larios, F.

2006-01-01

We give an overview of the physics of the Top quark, from the experimental discovery to the studies of its properties. We review some of the work done on the Electroweak and Flavor Changing couplings associated with the Top quark in the Standard Model and beyond. We will focus on the specific contribution of phycisits working in Mexico and Mexican physicists working abroad

Telekommunikatsiooni & arvutitootjate TOP

Index Scriptorium Estoniae

2006-01-01

Telekommunikatsiooni TOP 30. Tabelid. Vt. samas: Indrek Kald. Elisa ka tänavu esikohal; Väike Tele2 võidab paindlikkusest; Kernel läks üle piiri. Diagrammid. Arvutifirmade TOP 100. Tabelid. Vt. samas: Enn Heinsoo. Esikoha tõi kontori müük; Proeksperdil suund välisturgudele;¡ Ida-Eesti turg arenevale firmale kitsas. Diagrammid. Nimekiri: Omanikud

Toiduainetööstuse TOP 95 aastal 2003

Index Scriptorium Estoniae

2004-01-01

Toiduainetööstuse TOP 95; Kasumi TOP 40; Käibe TOP 40; Kasumi kasvu TOP 30; Rentaabluse TOP 30; Alkoholitootjate TOP; Pagarite TOP; Käibe kasvu TOP 30; Omakapitali tootlikkuse TOP 30; Toiduainetööstuste üld- ja finantsandmed

Forward Top Physics at LHCb

CERN Multimedia

CERN. Geneva

2018-01-01

The first Run 2 measurement of top pair production in the dilepton channel at 13 TeV will be presented, along with previous Run 1 measurements in final states accessible to both single top and top pair production. Heavy flavour tagging strategies at LHCb will also be discussed.

The Jules Horowitz Reactor (JHR), a European Material Testing Reactor (MTR), with extended experimental capabilities

International Nuclear Information System (INIS)

Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

2003-01-01

The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation. To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10 14 ncm -2 s -1 and a fast flux of 6,4.10 14 ncm -2 s -1 , it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = deplacement per atom). The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (author)

Wear Analysis of Top Piston Ring to Reduce Top Ring Reversal Bore Wear

Directory of Open Access Journals (Sweden)

P. Ilanthirayan

2017-12-01

Full Text Available The piston rings are the most important part in engine which controls the lubricating oil consumption and blowby of the gases. The lubricating film of oil is provided to seal of gases towards crankcase and also to give smooth friction free translatory motion between rings and liner. Of the three rings present top ring is more crucial as it does the main work of restricting gases downwards the crankcase. Boundary lubrication is present at the Top dead centre (TDC and Bottom dead centre (BDC of the liner surface. In addition to this, top ring is exposed to high temperature gases which makes the oil present near the top ring to get evaporated and decreasing its viscosity, making metal-metal contact most of the time. Due to this at TDC, excess wear happens on the liner which is termed as Top ring reversal bore wear. The wear rate depends upon many parameters such as lubrication condition, viscosity index, contact type, normal forces acting on ring, geometry of ring face, surface roughness, material property. The present work explores the wear depth for different geometries of barrel ring using Finite Element model with the help of Archard wear law and the same is validated through experimentation. The study reveals that Asymmetric barrel rings have less contact pressure which in turn reduces the wear at Top dead centre.

Boosted top: experimental tools overview

CERN Document Server

Usai, Emanuele

2015-01-01

An overview of tools and methods for the reconstruction of high-boost top quark decays at the LHC is given in this report. The focus is on hadronic decays, in particular an overview of the current status of top quark taggers in physics analyses is presented. The most widely used jet substructure techniques, normally used in combination with top quark taggers, are reviewed. Special techniques to treat pileup in large cone jets are described, along with a comparison of the performance of several boosted top quark reconstruction techniques.

Could stops lighten the top?

International Nuclear Information System (INIS)

Bilal, A.; Ellis, J.; Fogli, G.L.

1990-01-01

The analysis of the presently available electroweak data including radiative corrections in the standard model suggests that the top quark weighs more than the Z 0 . We examine whether squark loops in the minimal supersymmetric model, particularly those involving stops (partners of the top quark), could reduce substantially the preferred range of top quark masses. Given the present lower bounds on squark masses, we find that stop effects can reduce the central value of m t by at most a few GeV, although they do make a very heavy top quark increasingly unlikely. (orig.)

Life management for a non replaceable structure: the reactor building

International Nuclear Information System (INIS)

Torres, V.; Francia, L.

1998-01-01

Phase 1 of UNESA N.P.P. Lifetime Management Project identified and ranked important components, relative to plant life management. The list showed the Reactor Containment Structure in the third position, and thirteen concrete structures were among the list top twenty. Since the Reactor Containment Building, together with the Reactor Vessel, is the only non-replaceable plant component, and has a big impact on the plant technical life, there is an increasing interest on understanding its behavior to maintain structural integrity. This paper presents: a) IAEA (International Atomic Energy Agency) Coordinated Research Program experiences and studies. Under this Program, international experts address the most frequent degradation mechanisms affecting the containment building. b) IAEA Aging Management Program adapted to our plants. The paper addresses the aging mechanisms affecting the concrete structures, reinforcing steel and prestress systems as well as the aging management programs and the mitigation and control methods. Finally, this paper presents a new module called STRUCTURES, included in phase 2 of the above mentioned project, which will monitor and document the different aging mechanisms and management programs described in item b) regarding the Reactor Containment Building (concrete liner, post stressing system, anchor elements). This module will also support the Maintenance Rule related practices. (Author)

Measurement of the Mass Difference Between Top and Anti-top Quarks at CDF

CERN Document Server

Aaltonen, T.; Amerio, S.; Amidei, D.; Anastassov, A.; Annovi, A.; Antos, J.; Apollinari, G.; Appel, J.A.; Arisawa, T.; Artikov, A.; Asaadi, J.; Ashmanskas, W.; Auerbach, B.; Aurisano, A.; Azfar, F.; Badgett, W.; Bae, T.; Barbaro-Galtieri, A.; Barnes, V.E.; Barnett, B.A.; Barria, P.; Bartos, P.; Bauce, M.; Bedeschi, F.; Behari, S.; Bellettini, G.; Bellinger, J.; Benjamin, D.; Beretvas, A.; Bhatti, A.; Bisello, D.; Bizjak, I.; Bland, K.R.; Blumenfeld, B.; Bocci, A.; Bodek, A.; Bortoletto, D.; Boudreau, J.; Boveia, A.; Brigliadori, L.; Bromberg, C.; Brucken, E.; Budagov, J.; Budd, H.S.; Burkett, K.; Busetto, G.; Bussey, P.; Buzatu, A.; Calamba, A.; Calancha, C.; Camarda, S.; Campanelli, M.; Campbell, M.; Canelli, F.; Carls, B.; Carlsmith, D.; Carosi, R.; Carrillo, S.; Carron, S.; Casal, B.; Casarsa, M.; Castro, A.; Catastini, P.; Cauz, D.; Cavaliere, V.; Cavalli-Sforza, M.; Cerri, A.; Cerrito, L.; Chen, Y.C.; Chertok, M.; Chiarelli, G.; Chlachidze, G.; Chlebana, F.; Cho, K.; Chokheli, D.; Chung, W.H.; Chung, Y.S.; Ciocci, M.A.; Clark, A.; Clarke, C.; Compostella, G.; Convery, M.E.; Conway, J.; Corbo, M.; Cordelli, M.; Cox, C.A.; Cox, D.J.; Crescioli, F.; Cuevas, J.; Culbertson, R.; Dagenhart, D.; d'Ascenzo, N.; Datta, M.; de Barbaro, P.; Dell'Orso, M.; Demortier, L.; Deninno, M.; Devoto, F.; d'Errico, M.; Di Canto, A.; Di Ruzza, B.; Dittmann, J.R.; D'Onofrio, M.; Donati, S.; Dong, P.; Dorigo, M.; Dorigo, T.; Ebina, K.; Elagin, A.; Eppig, A.; Erbacher, R.; Errede, S.; Ershaidat, N.; Eusebi, R.; Farrington, S.; Feindt, M.; Fernandez, J.P.; Field, R.; Flanagan, G.; Forrest, R.; Frank, M.J.; Franklin, M.; Freeman, J.C.; Funakoshi, Y.; Furic, I.; Gallinaro, M.; Garcia, J.E.; Garfinkel, A.F.; Garosi, P.; Gerberich, H.; Gerchtein, E.; Giagu, S.; Giakoumopoulou, V.; Giannetti, P.; Gibson, K.; Ginsburg, C.M.; Giokaris, N.; Giromini, P.; Giurgiu, G.; Glagolev, V.; Glenzinski, D.; Gold, M.; Goldin, D.; Goldschmidt, N.; Golossanov, A.; Gomez, G.; Gomez-Ceballos, G.; Goncharov, M.; Gonzalez, O.; Gorelov, I.; Goshaw, A.T.; Goulianos, K.; Grinstein, S.; Grosso-Pilcher, C.; Group, R.C.; Guimaraes da Costa, J.; Hahn, S.R.; Halkiadakis, E.; Hamaguchi, A.; Han, J.Y.; Happacher, F.; Hara, K.; Hare, D.; Hare, M.; Harr, R.F.; Hatakeyama, K.; Hays, C.; Heck, M.; Heinrich, J.; Herndon, M.; Hewamanage, S.; Hocker, A.; Hopkins, W.; Horn, D.; Hou, S.; Hughes, R.E.; Hurwitz, M.; Husemann, U.; Hussain, N.; Hussein, M.; Huston, J.; Introzzi, G.; Iori, M.; Ivanov, A.; James, E.; Jang, D.; Jayatilaka, B.; Jeon, E.J.; Jindariani, S.; Jones, M.; Joo, K.K.; Jun, S.Y.; Junk, T.R.; Kamon, T.; Karchin, P.E.; Kasmi, A.; Kato, Y.; Ketchum, W.; Keung, J.; Khotilovich, V.; Kilminster, B.; Kim, D.H.; Kim, H.S.; Kim, J.E.; Kim, M.J.; Kim, S.B.; Kim, S.H.; Kim, Y.K.; Kim, Y.J.; Kimura, N.; Kirby, M.; Klimenko, S.; Knoepfel, K.; Kondo, K.; Kong, D.J.; Konigsberg, J.; Kotwal, A.V.; Kreps, M.; Kroll, J.; Krop, D.; Kruse, M.; Krutelyov, V.; Kuhr, T.; Kurata, M.; Kwang, S.; Laasanen, A.T.; Lami, S.; Lammel, S.; Lancaster, M.; Lander, R.L.; Lannon, K.; Lath, A.; Latino, G.; LeCompte, T.; Lee, E.; Lee, H.S.; Lee, J.S.; Lee, S.W.; Leo, S.; Leone, S.; Lewis, J.D.; Limosani, A.; Lin, C.J.; Lindgren, M.; Lipeles, E.; Lister, A.; Litvintsev, D.O.; Liu, C.; Liu, H.; Liu, Q.; Liu, T.; Lockwitz, S.; Loginov, A.; Lucchesi, D.; Lueck, J.; Lujan, P.; Lukens, P.; Lungu, G.; Lys, J.; Lysak, R.; Madrak, R.; Maeshima, K.; Maestro, P.; Malik, S.; Manca, G.; Manousakis-Katsikakis, A.; Margaroli, F.; Marino, C.; Martinez, M.; Mastrandrea, P.; Matera, K.; Mattson, M.E.; Mazzacane, A.; Mazzanti, P.; McFarland, K.S.; McIntyre, P.; McNulty, R.; Mehta, A.; Mehtala, P.; Mesropian, C.; Miao, T.; Mietlicki, D.; Mitra, A.; Miyake, H.; Moed, S.; Moggi, N.; Mondragon, M.N.; Moon, C.S.; Moore, R.; Morello, M.J.; Morlock, J.; Movilla Fernandez, P.; Mukherjee, A.; Muller, Th.; Murat, P.; Mussini, M.; Nachtman, J.; Nagai, Y.; Naganoma, J.; Nakano, I.; Napier, A.; Nett, J.; Neu, C.; Neubauer, M.S.; Nielsen, J.; Nodulman, L.; Noh, S.Y.; Norniella, O.; Oakes, L.; Oh, S.H.; Oh, Y.D.; Oksuzian, I.; Okusawa, T.; Orava, R.; Ortolan, L.; Pagan Griso, S.; Pagliarone, C.; Palencia, E.; Papadimitriou, V.; Paramonov, A.A.; Patrick, J.; Pauletta, G.; Paulini, M.; Paus, C.; Pellett, D.E.; Penzo, A.; Phillips, T.J.; Piacentino, G.; Pianori, E.; Pilot, J.; Pitts, K.; Plager, C.; Pondrom, L.; Poprocki, S.; Potamianos, K.; Prokoshin, F.; Pranko, A.; Ptohos, F.; Punzi, G.; Rahaman, A.; Ramakrishnan, V.; Ranjan, N.; Redondo, I.; Renton, P.; Rescigno, M.; Riddick, T.; Rimondi, F.; Ristori, L.; Robson, A.; Rodrigo, T.; Rodriguez, T.; Rogers, E.; Rolli, S.; Roser, R.; Ruffini, F.; Ruiz, A.; Russ, J.; Rusu, V.; Safonov, A.; Sakumoto, W.K.; Sakurai, Y.; Santi, L.; Sato, K.; Saveliev, V.; Savoy-Navarro, A.; Schlabach, P.; Schmidt, A.; Schmidt, E.E.; Schwarz, T.; Scodellaro, L.; Scribano, A.; Scuri, F.; Seidel, S.; Seiya, Y.; Semenov, A.; Sforza, F.; Shalhout, S.Z.; Shears, T.; Shepard, P.F.; Shimojima, M.; Shochet, M.; Shreyber-Tecker, I.; Simonenko, A.; Sinervo, P.; Sliwa, K.; Smith, J.R.; Snider, F.D.; Soha, A.; Sorin, V.; Song, H.; Squillacioti, P.; Stancari, M.; St. Denis, R.; Stelzer, B.; Stelzer-Chilton, O.; Stentz, D.; Strologas, J.; Strycker, G.L.; Sudo, Y.; Sukhanov, A.; Suslov, I.; Takemasa, K.; Takeuchi, Y.; Tang, J.; Tecchio, M.; Teng, P.K.; Thom, J.; Thome, J.; Thompson, G.A.; Thomson, E.; Toback, D.; Tokar, S.; Tollefson, K.; Tomura, T.; Tonelli, D.; Torre, S.; Torretta, D.; Totaro, P.; Trovato, M.; Ukegawa, F.; Uozumi, S.; Varganov, A.; Vazquez, F.; Velev, G.; Vellidis, C.; Vidal, M.; Vila, I.; Vilar, R.; Vizan, J.; Vogel, M.; Volpi, G.; Wagner, P.; Wagner, R.L.; Wakisaka, T.; Wallny, R.; Wang, S.M.; Warburton, A.; Waters, D.; Wester, W.C., III; Whiteson, D.; Wicklund, A.B.; Wicklund, E.; Wilbur, S.; Wick, F.; Williams, H.H.; Wilson, J.S.; Wilson, P.; Winer, B.L.; Wittich, P.; Wolbers, S.; Wolfe, H.; Wright, T.; Wu, X.; Wu, Z.; Yamamoto, K.; Yamato, D.; Yang, T.; Yang, U.K.; Yang, Y.C.; Yao, W.M.; Yeh, G.P.; Yi, K.; Yoh, J.; Yorita, K.; Yoshida, T.; Yu, G.B.; Yu, I.; Yu, S.S.; Yun, J.C.; Zanetti, A.; Zeng, Y.; Zhou, C.; Zucchelli, S.

2013-03-18

We present a measurement of the mass difference between top ($t$) and anti-top ($\\bar{t}$) quarks using $t\\bar{t}$ candidate events reconstructed in the final state with one lepton and multiple jets. We use the full data set of Tevatron $\\sqrt{s} = 1.96$ TeV proton-antiproton collisions recorded by the CDF II detector, corresponding to an integrated luminosity of 8.7 fb$^{-1}$. We estimate event-by-event the mass difference to construct templates for top-quark signal events and background events. The resulting mass difference distribution of data compared to signal and background templates using a likelihood fit yields $\\Delta M_{top} = {M}_{t} - {M}_{\\bar{t}} = -1.95 $pm$ 1.11 (stat) $pm$ 0.59 (syst)$ and is in agreement with the standard model prediction of no mass difference.

A robot-automated work site for repair of the Chinon A3 reactor

International Nuclear Information System (INIS)

Raynal, A.

1987-01-01

In 1982, following degradation due to corrosion of low-carbon steel by carbon dioxide gas, the utility undertook to repair some of the support structures at Chinon A3. This involved consolidation and reinforcing thermocouples and gas monitor pipeworks supports. A welding process was selected and the use of robots became indispensable because of the large number of components to be replaced (200 per outage). Two robots, supplied with tool heads and replacement components from outside the reactor were used. The robots and their servers were coordinated by a central computer and monitored by a closed circuit television system. Each repair operation was performed after ''training'' on a full-scale mockup of the top of the reactor reconstructed from telemetry of the real reactor dimensions. Since becoming operational in June 1986, the robots have accumulated over 20 000 hours of operation and seventy parts have been welded to the reactor. A 3D CAD system has been adapted to simulate the robots and analyse long trajectories in order to reduce robot learning time [fr

Local Physics Basis of Confinement Degradation in JET ELMy H-Mode Plasmas and Implications for Tokamak Reactors

International Nuclear Information System (INIS)

Budny, R.V.; Alper, B.; Borba, D.; Cordey, J.G.; Ernst, D.R.; Gowers, C.

2001-01-01

First results of gyrokinetic analysis of JET [Joint European Torus] ELMy [Edge Localized Modes] H-mode [high-confinement modes] plasmas are presented. ELMy H-mode plasmas form the basis of conservative performance predictions for tokamak reactors of the size of ITER [International Thermonuclear Experimental Reactor]. Relatively high performance for long duration has been achieved and the scaling appears to be favorable. It will be necessary to sustain low Z(subscript eff) and high density for high fusion yield. This paper studies the degradation in confinement and increase in the anomalous heat transport observed in two JET plasmas: one with an intense gas puff and the other with a spontaneous transition between Type I to III ELMs at the heating power threshold. Linear gyrokinetic analysis gives the growth rate, gamma(subscript lin) of the fastest growing modes. The flow-shearing rate omega(subscript ExB) and gamma(subscript lin) are large near the top of the pedestal. Their ratio decreases approximately when the confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently high |gamma(subscript ExB)|/gamma(subscript lin) near the top of the pedestal for high confinement

Toiduainetööstuse TOP 95 aastal 2002

Index Scriptorium Estoniae

2003-01-01

Toiduainetööstuse TOP 95. Käibe TOP 50. Kasumi TOP 50. Rentaabluse TOP 25. Kasumi kasvu TOP 25. Omakapitali tootlikkuse TOP 25. Käibe kasvu TOP 25. Toiduainetööstuse TOP 95 ettevõtted 2002: üldandmed ja finantsandmed

Top physics in WHIZARD

Energy Technology Data Exchange (ETDEWEB)

Reuter, Juergen; Chokoufe Nejad, Bijan [DESY, Hamburg (Germany). Theory Group; Bach, Fabian [European Commission, Eurostat, Luxembourg (Luxembourg); Hoang, Andre [Vienna Univ. (Austria). Faculty of Physics; Vienna Univ. (Austria). Erwin Schroedinger International Inst. for Mathematical Physics; Kilian, Wolfgang [Siegen Univ. (Germany); Stahlhofen, Maximilian [Mainz Univ. (Germany). PRISMA Cluster of Excellence; Teubner, Thomas [Liverpool Univ. (United Kingdom). Dept. of Mathematical Sciences; Weiss, Christian [DESY, Hamburg (Germany). Theory Group; Siegen Univ. (Germany)

2016-03-15

In this talk we summarize the top physics setup in the event generator WHIZARD with a main focus on lepton colliders. This includes full six-, eight- and ten-fermion processes, factorized processes and spin correlations. For lepton colliders, QCD NLO processes for top quark physics are available and discussed. A special focus is on the top-quark pair threshold, where a special implementation combines a non-relativistic effective field theory calculation augmented by a next-to-leading threshold logarithm resummation with a continuum relativistic fixed-order QCD NLO simulation.

Top physics in WHIZARD

International Nuclear Information System (INIS)

Reuter, Juergen; Chokoufe Nejad, Bijan; Hoang, Andre; Stahlhofen, Maximilian

2016-03-01

In this talk we summarize the top physics setup in the event generator WHIZARD with a main focus on lepton colliders. This includes full six-, eight- and ten-fermion processes, factorized processes and spin correlations. For lepton colliders, QCD NLO processes for top quark physics are available and discussed. A special focus is on the top-quark pair threshold, where a special implementation combines a non-relativistic effective field theory calculation augmented by a next-to-leading threshold logarithm resummation with a continuum relativistic fixed-order QCD NLO simulation.

Top production at hadron colliders

Indian Academy of Sciences (India)

New results on top quark production are presented from four hadron collider experiments: CDF and D0 at the Tevatron, and ATLAS and CMS at the LHC. Cross-sections for single top and top pair production are discussed, as well as results on the top–antitop production asymmetry and searches for new physics including ...

Top Production at LHCb

CERN Multimedia

Santana Rangel, Murilo

2015-01-01

Single and pair top production in the forward direction at the LHC allows for precision tests of the Standard Model. The observation of top quarks in 7 and 8 TeV data and prospects for precision measurements are shown.

Search for Top in TOTEM

CERN Document Server

Corrales, Alonso

2017-01-01

The work assigned to me during this summer (from June 19th to August 11th) was in the "Search for Top in TOTEM" project, a CMG project in collaboration with TOTEM, under the supervision of Martijn Mulders and Laurent Forthomme. The goal is to determine physical properties of the top anti top quark pair (tt) created by central exclusive processes, particularly the invariant mass. For that it is necessary to relate the measurements of the reconstructed top pair kinematics in the central detector of CMS with the detected protons in the Roman Pots of TOTEM just outside the interaction site, about 200 m from it. CMS only considers decay products from the proton collision, it does not regard the forward protons that participated in the collision. On the contrary, TOTEM does consider these protons. In this research measurements of both TOTEM and CMS were combined Figure 1 shows a diagram of the production of the top anti top pair from the interaction of the two protons. A central exclusive process occurs when the pr...

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Architecture of top down, parallel pattern recognition system TOPS and its application to the MR head images

International Nuclear Information System (INIS)

Matsunoshita, Jun-ichi; Akamatsu, Shigeo; Yamamoto, Shinji.

1993-01-01

This paper describes about the system architecture of a new image recognition system TOPS (top-down parallel pattern recognition system), and its application to the automatic extraction of brain organs (cerebrum, cerebellum, brain stem) from 3D-MRI images. Main concepts of TOPS are as follows: (1) TOPS is the top-down type recognition system, which allows parallel models in each level of hierarchy structure. (2) TOPS allows parallel image processing algorithms for one purpose (for example, for extraction of one special organ). This results in multiple candidates for one purpose, and judgment to get unique solution for it will be made at upper level of hierarchy structure. (author)

Measurements and searches with top quarks

International Nuclear Information System (INIS)

Peters, Reinhild Yvonne

2008-01-01

In 1995 the last missing member of the known families of quarks, the top quark, was discovered by the CDF and D0 experiments at the Tevatron, a proton-antiproton collider at Fermilab near Chicago. Until today, the Tevatron is the only place where top quarks can be produced. The determination of top quark production and properties is crucial to understand the Standard Model of particle physics and beyond. The most striking property of the top quark is its mass--of the order of the mass of a gold atom and close to the electroweak scale--making the top quark not only interesting in itself but also as a window to new physics. Due to the high mass, much higher than of any other known fermion, it is expected that the top quark plays an important role in electroweak symmetry breaking, which is the most prominent candidate to explain the mass of particles. In the Standard Model, electroweak symmetry breaking is induced by one Higgs field, producing one additional physical particle, the Higgs boson. Although various searches have been performed, for example at the Large Electron Positron Collider (LEP), no evidence for the Higgs boson could yet be found in any experiment. At the Tevatron, multiple searches for the last missing particle of the Standard Model are ongoing with ever higher statistics and improved analysis techniques. The exclusion or verification of the Higgs boson can only be achieved by combining many techniques and many final states and production mechanisms. As part of this thesis, the search for Higgs bosons produced in association with a top quark pair (t(bar t)H) has been performed. This channel is especially interesting for the understanding of the coupling between Higgs and the top quark. Even though the Standard Model Higgs boson is an attractive candidate, there is no reason to believe that the electroweak symmetry breaking is induced by only one Higgs field. In many models more than one Higgs boson are expected to exist, opening even more channels

Measurement of top properties in ATLAS

CERN Document Server

Veloso, F; The ATLAS collaboration

2010-01-01

The top quark may play a special role in the Standard Model of particle physics. It is, for instance, the heaviest fundamental particle known, with a mass close to the electroweak symmetry breaking scale. Millions of top quark pairs will be produced per year at the LHC at nominal luminosity. ATLAS will measure many of the top quark properties with unprecedented precision. The ATLAS sensitivity studies done with Monte Carlo for the measurements of the top quark mass, the top quark charge, rare top quark decays and flavour changing neutral currents, the $t\\bar t$ spin correlations and $W$ boson polarization, the anomalous couplings at the $Wtb$ vertex and $t\\bar t$ resonances are reviewed.

Top quark mass measurements with CMS

CERN Document Server

Kovalchuk, Nataliia

2017-01-01

Measurements of the top quark mass are presented, obtained from CMS data collected in proton-proton collisions at the LHC at centre-of-mass energies of 7 TeV and 8 TeV. The mass of the top quark is measured using several methods and channels, including the reconstructed invariant mass distribution of the top quark, an analysis of endpoint spectra as well as measurements from shapes of top quark decay distributions. The dependence of the mass measurement on the kinematic phase space is investigated. The results of the various channels are combined and compared to the world average. The top mass and also $\\alpha_{\\textnormal S}$ are extracted from the top pair cross section measured at CMS.

Top quark physics at the LHC

Directory of Open Access Journals (Sweden)

Jeong Kim Tae

2014-04-01

Full Text Available In 2011, an integrated luminosity of more than 5 fb−1 at 7 TeV has been delivered by the LHC. The measurement of the cross section in top quark pair production and in single top quark production, top quark mass, top quark properties and new physics searches in top quark decays have been performed at the CMS experiment with various integrated luminosities. An overview of the latest results of these measurements and searches by the time of ICFP 2012 conference will be presented.

Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II; Desgaste placas tubos guia barras de control interno superior vasija del reactor C.N. Vandellos II

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-07-01

The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

Thermal barrier and support for nuclear reactor fuel core

International Nuclear Information System (INIS)

Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

1987-01-01

A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

Top Physics at CMS/LHC

Directory of Open Access Journals (Sweden)

Daskalakis Georgios

2017-01-01

Full Text Available Recent results on the inclusive and differential production cross sections of top-quark pair and single top-quark processes are presented, obtained using data from proton-proton collisions collected with the CMS detector at the LHC. The large centre-of-mass energies available at LHC allow for the copious production of top-quark pairs in association with other final state particles at high transverse momentum. Measurements of such processes as well as of the top-quark mass and other properties will be discussed. The results are compared with the most up-to-date standard model theory predictions.

Development of special tools for the cleaning of reactor's interior in HANARO

International Nuclear Information System (INIS)

Cho, Y.-G.; Le, J.-H.; Ryu, J.-S.; Wu, J.-S.; Jung, H.-S.

1999-01-01

The HANARO (Hi-flux Advanced Neutron Application Reactor) in Korea has been being operated for 5 years, including one year of non-nuclear system commissioning tests since the installation of the reactor in early 1994. The HANARO is an open-tank-in-pool type reactor which has the advantage of free access from the pool top. The HANARO reactor had special cleaning works twice to remove debris from the inside reactor. This paper summarizes the development of special tools for reactor cleaning and how the reactor's inside had been successfully cleaned within short periods. The first cleaning work, after the initial flushing of the reactor system in early 1994, was the removal of the silica-gel sands, contaminated during installation, from the reactor pool and all equipment in the pool, including the reactor structure, the reactivity control units and the primary cooling system. Water-jet, pump suction, vacuum suction and whirl methods were used in combination with specially designed tools. The second one, occurred in February 1997 after two years of reactor operation was the cleaning work for the reactor's interior to remove several metal pieces broken from the parts of a check valve assembly in the primary cooling system. This work required development of many special tools that are all compact in size and remotely operable to reach all areas of the inlet plenum through very limited access holes. The special tools used for this work were two kinds of underwater cameras equipped with lighting, a debris-picking tool named 'revolving dustpan', two kinds of flow tube replacement tools and many other supplementary tools. All work had been successfully accomplished on the in-pool-platform temporarily installed 9m above the pool bottom to maintain the pool water level required in view of radiation shielding. Finally, the reactor internals were inspected using the underwater cameras to confirm the absence of debris and the surface integrity of the plenum as well as all fuel

Lääne-Virumaa TOP 100 aastal 2000

Index Scriptorium Estoniae

2001-01-01

Lääne-Virumaa edukamad ettevõtted; Lääne-Virumaa käibe TOP 100; Käibe kasvu TOP 20; Käibe languse TOP 10; Kasumi TOP 20; Kasumi kasvu TOP 20; Rentabluse TOP 20; ROA TOP 20; Kasumi languse TOP 10; Kahjumi TOP 10; Lääne-Virumaa käibelt suuremate ettevõtete finantsandmed. Lääne-Virumaa ettevõtete üldandmed

Measurements of Top Properties at the Tevatron

International Nuclear Information System (INIS)

Husemann, Ulrich; Yale U.

2007-01-01

The large data samples of thousands of top events collected at the Tevatron experiments CDF and D(O) allow for a variety of measurements to analyze the properties of the top quark. Guided by the question ''Is the top quark observed at the Tevatron really the top quark of the standard model,'' we present Tevatron analyses studying the top production mechanism including resonant t(bar t) production, the V -A structure of the t → Wb decay vertex, the charge of the top quark, and single-top production via flavor-changing neutral currents

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Rõivatööstuse TOP 40 aastal 2002

Index Scriptorium Estoniae

2003-01-01

Rõivatööstuse TOP 40. Käibe TOP 40. Kasumi TOP 40. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Rõivatööstuse ettevõtted: üldandmed ja finantsandmed

Ravimimüüjate TOP 30 aastal 2002

Index Scriptorium Estoniae

2003-01-01

Ravimimüüjate TOP 30 aastal 2002. Käibe TOP 30. Kasumi TOP 30. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Rentaabluse TOP 20. Omakapitali tootlikkuse TOP 20. Ravimimüüjad 2002: üld- ja finantsandmed

Tekstiilitööstuse TOP 30 aastal 2002

Index Scriptorium Estoniae

2003-01-01

Tekstiilitööstuse TOP 30. Kasumi TOP 30. Käibe TOP 30. Rentaabluse TOP 20. Käibe kasvu TOP 20. Kasumi kasvu TOP 20. Omakapitali tootlikkuse TOP 20. Tekstiilitööstuse ettevõtted 2002: üld- ja finantsandmed

Influence of partial blockage of a BWR bundle on heat transfer, cladding temperature, and quenching during bottom flooding or top spraying under simulated LOCA conditions

International Nuclear Information System (INIS)

Brand, B.; Gaul, H.P.; Sarkar, J.

1982-01-01

In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m 3 /h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800 0 C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis

5 CFR 1312.27 - Top secret control.

Science.gov (United States)

2010-01-01

... 5 Administrative Personnel 3 2010-01-01 2010-01-01 false Top secret control. 1312.27 Section 1312... Classified Information § 1312.27 Top secret control. The EOP Security Officer serves as the Top Secret... Top Secret material. The ATSCOs will be responsible for the accountability and custodianship of Top...

Lepton distribution in top decay: A probe of new physics and top ...

Indian Academy of Sciences (India)

it may become difficult to probe new physics couplings of top quark. However, if one can construct observables that are sensitive to production and decay mecha- nism independent of each other, the analysis can be greatly simplified. It has been shown that the angular distribution of the leptons from the decay of top quark is.

Toiduainetööstuse TOP 100 aastal 2001

Index Scriptorium Estoniae

2002-01-01

TOP 100. Toiduainetööstuste rentaabluse TOP 40. Varade tootlikkuse TOP 40. Käibe TOP 50. Käibe kasvu TOP 50. Kasumi TOP 50. Kasumi kasvu TOP 50. Toiduainetööstuse finantsandmed aastal 2001. Toiduainetööstuse üldandmed aastal 2001

Trükitööstuste TOP 50

Index Scriptorium Estoniae

2002-01-01

Trükitööstuste TOP 50. Käibe TOP 30. Käibe kasum TOP 30. Kasvu TOP 30. Kasumi kasvu TOP 30. Rentaabluse TOP. Varade tootlikkuse TOP. Trükitööstusettevõtete üldandmed. Trükitööstusettevõtete finantsandmed

Method of detecting stacks with leaky fuel elements in liquid-metal-cooled reactor and apparatus for effecting same

International Nuclear Information System (INIS)

Aristarkhov, N.N.; Efimov, I.A.; Zaistev, B.I.; Peters, I.G.; Tymosh, B.S.

1976-01-01

Described is a method of detecting stacks with leaky fuel elements in a liquid-metal-cooled reactor, consisting in that prior to withdrawing a coolant sample, gas is accumulated in the coolant of the stack being controlled, the reactor being shut down, separated from the sample by means of an inert carrier gas, and the radioactivity of the separated gas is measured. An apparatus for carrying out said method comprises a sampler in the form of a tube parallel to the reactor axis in the hole of a rotating plug and adapted to move along the reactor axis. Made in the top portion of the tube are holes for the introduction of the inert carrier gas and the removal thereof together with the gases evolved from the coolant, while the bottom portion of the tube is provided with a sealing member

New Physics in Single-Top Production

CERN Document Server

Kind, OM; The ATLAS collaboration

2013-01-01

In this presentation for TOP 2013 the latest results on searches of physics beyond the Standard Model using single-top signatures from CDF, CMS, D0 and ATLAS are collected. This includes searches for unknown resonances like W' or b*, measurements of the W helicity fractions and top polarisation in single-top events, as well as tests for CP violation, FCNC or anomalous weak couplings.

Discovery of single top quark production

Energy Technology Data Exchange (ETDEWEB)

Gillberg, Dag [Simon Fraser Univ., Burnaby, BC (Canada)

2009-04-01

The top quark is by far the heaviest known fundamental particle with a mass nearing that of a gold atom. Because of this strikingly high mass, the top quark has several unique properties and might play an important role in electroweak symmetry breaking - the mechanism that gives all elementary particles mass. Creating top quarks requires access to very high energy collisions, and at present only the Tevatron collider at Fermilab is capable of reaching these energies. Until now, top quarks have only been observed produced in pairs via the strong interaction. At hadron colliders, it should also be possible to produce single top quarks via the electroweak interaction. Studies of single top quark production provide opportunities to measure the top quark spin, how top quarks mix with other quarks, and to look for new physics beyond the standard model. Because of these interesting properties, scientists have been looking for single top quarks for more than 15 years. This thesis presents the first discovery of single top quark production. An analysis is performed using 2.3 fb^-1 of data recorded by the D0 detector at the Fermilab Tevatron Collider at centre-of-mass energy √s = 1.96 TeV. Boosted decision trees are used to isolate the single top signal from background, and the single top cross section is measured to be σ(p$\\bar{p}$ → tb + X, tqb + X) = 3.74_-0.74^+0.95 pb. Using the same analysis, a measurement of the amplitude of the CKM matrix element V_tb, governing how top and b quarks mix, is also performed. The measurement yields: |V{sub tb}|f₁^L| = 1.05 -_0.12^+0.13, where f₁^L is the left-handed Wtb coupling. The separation of signal from background is improved by combining the boosted decision trees with two other multivariate techniques. A new cross section measurement is performed, and the significance for the excess over the predicted background exceeds 5

Single top quark production with CMS

Directory of Open Access Journals (Sweden)

Piccolo Davide

2013-11-01

Full Text Available Measurements of single top quark production performed using the CMS experiment [1] data collected in 2011 at centre-of-mass energies of 7 TeV and in 2012 at 8 TeV, are presented. The cross sections for the electroweak production of single top quarks in the t-channel and in association with W-bosons is measured and the results are used to place constraints on the CKM matrix element Vtb. Measurements of top quark properties in single top quark production are also presented. The results include the measurement of the charge ratio in the single top t-channel.

The Bottom supported fast breeder reactor vessel - an alternative approach to seismic accommodation and reduced cost

International Nuclear Information System (INIS)

Nakagawa, H.; Golan, S.; Petrozelli, J.; Kumaoka, Y.; Kawamura, Y.

1988-01-01

Most FBR vessels are supported by hanging from their top portions. A disadvantage of such a top supported reactor vessel (TSRV) structural configuration is that it may generate high reactor core accelerations. This is due to the long path the seismic vibrations must travel from the basemat up through the building and then down through the RV block to the core. To compensate for this disadvantage, TSRV blocks are often strengthened beyond what is required for other considerations, such as pressure, to satisfy seismic response criteria, thus increasing weights and costs. In addition to long load paths, TSRVs also have common load paths. For example, in a TSRV (with the core supported from the bottom of the RV) the sodium and core loads both travel along the RV pressure boundary. Therefore, one of these loads will likely control the RV thickness leaving excess margin for the other loads. It is the premise of this paper that the revision of a large pool FBR from a TSRV configuration to a specific bottom supported reactor vessel (BSRV) configuration can resolve the above TSRV disadvantages related to load path length and diversity, thereby improving seismic performance and simultaneously reducing RV block costs by reducing weights. This paper demonstrates this premise by comparing a reference TSRV block with a specific BSRV block design

An Efficiency Comparison of MBA Programs: Top 10 versus Non-Top 10

Science.gov (United States)

Hsu, Maxwell K.; James, Marcia L.; Chao, Gary H.

2009-01-01

The authors compared the cohort group of the top-10 MBA programs in the United States with their lower-ranking counterparts on their value-added efficiency. The findings reveal that the top-10 MBA programs in the United States are associated with statistically higher average "technical and scale efficiency" and "scale efficiency", but not with a…

Air box shock absorber for a nuclear reactor

International Nuclear Information System (INIS)

Pradhan, A.V.; George, J.A.

1977-01-01

Disclosed is an air box type shock absorber primarily for use in an ice condenser compartment of a nuclear reactor. The shock absorber includes a back plate member and sheet metal top, bottom, and front members. The front member is prefolded, and controlled clearances are provided among the members for predetermined escape of air under impact and compression. Prefolded internal sheet metal stiffeners also absorb a portion of the kinetic energy imparted to the shock absorber, and limit rebound. An external restraining rod guided by restraining straps insures that the sheet metal front member compresses inward upon impact. 6 claims, 11 figures

New directions in top physics

Energy Technology Data Exchange (ETDEWEB)

Schell, Torben Karl

2017-02-01

The top quark plays an important role for many aspects of particle physics. The coupling of the Higgs boson to top quarks is a key parameter to probe electroweak symmetry breaking and is important for the evolution of the Higgs potential to high energies. In addition, many models of physics beyond the Standard Model predict heavy particles that decay to top-quark pairs. Furthermore, the unexplained hierarchy of fermion masses culminates in the large top-quark mass. In this thesis, we consider resonance searches based on top quarks in the fully hadronic final state. We employ multivariate techniques in form of boosted decision trees and add several improvements to the original HEPTopTagger algorithm. These modifications and extensions result in the new HEPTopTagger2. The achieved improvements are used to estimate the precision to which the top Yukawa coupling can be measured at a future 100 TeV proton-proton collider in the fully hadronic final state of t anti tH production. We find that at such a collider a precision measurement of the top Yukawa coupling to 1% should be possible. The statistical precision is backed up by demonstrating that in the ratio σ(t anti tH)/σ(t anti tZ) theoretical uncertainties cancel to below-percent level. Finally, we propose a Froggatt-Nielsen-type model to address the hierarchy of fermion masses in the Standard Model and determine current and projected bounds on the available parameter space.

Top mass in ATLAS and CMS

CERN Document Server

Castro, A.

2017-01-01

Top quarks are produced copiously at the LHC, and a variety of related measurements has been made in the recent years by the two collaborations ATLAS and CMS. The most recent measurements of the top quark mass by the two collaborations are reported here. The top quark mass has been measured with a relative uncertainty smaller than 0.3pct, making the top quark the most accurately measured quark.

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Measurements of turbulence in a microscale multi-inlet vortex nanoprecipitation reactor

International Nuclear Information System (INIS)

Shi, Yanxiang; Cheng, Janine Chungyin; Fox, Rodney O; Olsen, Michael G

2013-01-01

The microscale multi-inlet vortex reactor (MIVR) is designed for use in Flash NanoPrecipitation (FNP), a promising technique for producing nanoparticles within small particle size distribution. Fluid mixing is crucial in the FNP process, and due to mixing’s strong dependence upon fluid kinematics, investigating velocity and turbulence within the reactor is crucial to optimizing reactor design. To this end, microscopic particle image velocimetry has been used to investigate flow within the MIVR. Three Reynolds numbers are studied, namely, Re j = 53, 93 and 240. At Re j = 53, the flow is laminar and steady. Due to the strong viscous effects at this Reynolds number, distinct flow patterns are observed at different distances from the reactor top and bottom walls. The viscous effects also retard the tangential motions within the reactor, resulting in a weaker vortex than appears at the higher Reynolds numbers. As the Reynolds number is increased to 93, the flow becomes more homogeneous over the depth of the reactor due to weaker viscous effects, yet the flow is still steady. The diminishing effects of viscosity also result in a stronger vortex. At the highest Reynolds number investigated, the flow is turbulent. Turbulent statistics including tangential and radial velocity fluctuations and Reynolds shear stresses are analyzed for this case in addition to the mean velocity field. The tangential motions of the flow are strongest at Re j = 240. Both the tangential and radial velocity fluctuations increase as the flow spirals toward the center of the reactor. The magnitudes of the tangential and radial velocity fluctuations are similar, suggesting that the turbulence is locally isotropic. (paper)

Device for providing a leak-tight penetration for electric cables through a reactor vault roof

International Nuclear Information System (INIS)

Eyral, M.; Mahe, A.

1979-01-01

The device for providing a cable penetration through the vault roof of a liquid sodium cooled fast reactor comprises a vertical tube closed at the top end by a flange-plate. Electric cables connected to measuring and detecting instruments are passed through the flange-plate which is joined to the reactor vault roof in leak-tight manner and enclosed within a removable hood. At least one horizontal plate is mounted within the vertical tube and provided with orifices for the leak-tight passage of the cables. Cable storage reels are placed within the tube and can be locked in position or released by controlled mechanical means

Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

International Nuclear Information System (INIS)

Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

2014-01-01

Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

International Nuclear Information System (INIS)

Samalova, Ludmila; Chvala, Ondrej; Maldonado, G. Ivan

2017-01-01

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

The theory of the top

CERN Document Server

Klein, Felix; Nagem, Raymond J; Sandri, Guido

The Theory of the Top. Volume IV. Technical Applications of the Theory of the Top is the fourth and final volume in a series of self-contained English translations of the classic and definitive treatment of rigid body motion. Key features: * Complete and unabridged presentation with recent advances and additional notes; * Annotations by the translators provide insights into the nature of science and mathematics in the late 19th century; * Each volume interweaves theory and applications. The Theory of the Top was originally presented by Felix Klein as an 1895 lecture at Göttingen University that was broadened in scope and clarified as a result of collaboration with Arnold Sommerfeld.  Graduate students and researchers interested in theoretical and applied mechanics will find this series of books a thorough and insightful account.  Other volumes in the series include Introduction to the Kinematics and Kinetics of the Top, Development of the Theory in the Case of the Heavy Symmetric Top, and Perturbation...

Toiduainetööstuse TOP 100 aastal 2000

Index Scriptorium Estoniae

2001-01-01

Toiduainetööstuse käibe TOP 75; Toiduainetööstuse käibekasvu TOP 60; Toiduainetööstuse kasumi TOP 80; Kahjumi TOP 15; Toiduainetööstuse kasumi kasvu TOP 60; Rentaabluse TOP 50; varade tootlikkuse TOP 50; Toiduainetööstuse finantsandmed aastal 2000; Toiduainetööstuse üldandmed aastal 2000

The Progress of Fast Reactor Technology Development in China

International Nuclear Information System (INIS)

Yang, Hongyi; Xu, Mi

1994-01-01

China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basis strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan(2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor(CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached the top namely 57m above the ground. More than one hundred components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000m 2 floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which in only under consideration up to now. Some important technical selections have been settled, but its design has not yet started

Ray Tracing Study on Top ECCD Launch in KSTAR

Directory of Open Access Journals (Sweden)

Bae Young-soon

2017-01-01

Full Text Available The current drive efficiency of electron cyclotron (EC wave is typically low compared with other RF and neutral beam heating system in tokamak. It is known that EC current drive by outboard launch suffers from low current drive efficiency due to electron trapping. However, the heating and current drive by EC wave is being regarded as a strong candidate for DEMO reactor due to the simplicity of the launcher, none of its interaction with plasma, and no coupling issue at the plasma edge. Also, off-axis heating and current drive by EC wave plays an important role of steady state operation optimization. To enhance the current drive efficiency in DEMO-relevant operation condition having high density and high temperature, the top launch of EC wave is recently proposed in FNSF design [2]. In FNSF, a top launch makes use of a large toroidal component to the launch direction adjusting the vertical launch angle so that the rays propagate nearly parallel to the resonance layer increasing of Doppler shift with higher n||. The results shows a high dimensional efficiency for a broad ECCD profile peaked off axis. In KSTAR, the possibility of efficient off-axis ECCD using top launch is investigated using the ray tracing code, GENRAY [3] for the operating EC frequencies (105 GHz or 140 GHz, and 170 GHz. The high current drive efficiency is found by adjusting the toroidal magnetic field and the radial pivot position of the final launcher mirror for fundamental O-mode and second harmonic X-mode. A large Doppler shift is not quite sure in the typical plasma profile in KSTAR, but the simulation results show high current drive efficiency. This paper presents ray tracing results for many cases with the wave trajectories and damping of EC by scanning the launching angle for specific launcher pivot positions and toroidal magnetic field, and two equilibriums of the KSTAR.

LIGHT WATER MODERATED NEUTRONIC REACTOR

Science.gov (United States)

Christy, R.F.; Weinberg, A.M.

1957-09-17

A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

Top wealth shares in Australia 1915-2012

OpenAIRE

Katic, Pamela; Leigh, Andrew

2016-01-01

Combining data from surveys, inheritance tax records, and rich lists, we estimate top wealth shares for Australia from World War I until the present day. We find that the top 1 percent share declined by two-thirds from 1915 until the late 1960s, and rose from the late 1970s to 2010. The recent increase is sharpest at the top of the distribution, with the top 0.001 percent wealth share tripling from 1984 to 2012. The trend in top wealth shares is similar to that in Australian top income shares...

Pebble-bed reactor

International Nuclear Information System (INIS)

Lohnert, G.; Mueller-Frank, U.; Heil, J.

1976-01-01

A pebble-bed nuclear reactor of large power rating comprises a container having a funnel-shaped bottom forming a pebble run-out having a centrally positioned outlet. A bed of downwardly-flowing substantially spherical nuclear fuel pebbles is positioned in the container and forms a reactive nuclear core maintained by feeding unused pebbles to the bed's top surface while used or burned-out pebbles run out and discharge through the outlet. A substantially conical body with its apex pointing upwardly and its periphery spaced from the periphery of the container spreads the bottom of the bed outwardly to provide an annular flow down the funnel-shaped bottom forming the runout, to the discharge outlet. This provides a largely constant downward velocity of the spheres throughout the diameter of the bed throughout a substantial portion of the down travel, so that all spheres reach about the same burned-out condition when they leave the core, after a single pass through the core area

SWFSC/MMTD/CCE: Tagging and Tracking of Physeter (T-TOP/T-TOP2) 1997

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This research was devoted to studying sperm whale (Physeter macrocephalus) behavior. The principal study area for T-TOP and T-TOP2 was the Pioneer Seamount, 50...

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

General outline of the operation and utilization of the BR2 reactor

International Nuclear Information System (INIS)

Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

1978-01-01

The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Top Management Involvement in New Product Development

DEFF Research Database (Denmark)

Felekoglu, Burcu; Maier, Anja; Moultrie, James

2010-01-01

a broader conceptual space than this participation. This paper reviews the literature on top management involvement in new product development (NPD) and discusses relevance of different theoretical perspectives from other disciplines such as management, organisational behaviour and communication to analyse......Involvement of top managers in new product development (NPD) is a critical factor affecting NPD performance and frequently considered to be the participation of top management to certain activities in NPD or their NPD related behaviours. However, “Top management involvement in NPD” occupies...... antecedents, realisation and consequences of top management involvement in NPD. It is argued that top management has different involvement at different NPD levels: organisation level and project level. Resulting from this literature review, a tentative framework for top management involvement in NPD...

Decommissioning of the AVR reactor, concept for the total dismantling

International Nuclear Information System (INIS)

Marnet, C.; Wimmers, M.; Birkhold, U.

1998-01-01

After more than 21 years of operation, the 15 MWe AVR experimental nuclear power plant with pebble bed high temperature gas-cooled reactor was shout down in 1988. Safestore decommissioning began in 1994. In order to completely dismantle the plant, a concept for Continued dismantling was developed according to which the plant could be dismantled in a step-wise procedure. After each step, there is the possibility to transform the plant into a new state of safe enclosure. The continued dismantling comprises three further steps following Safestore decommissioning: 1. Dismantling the reactor vessels with internals; 2. Dismantling the containment and the auxiliary units; 3. Gauging the buildings to radiation limit, release from the validity range of the AtG (Nuclear Act), and demolition of the buildings. For these steps, various technical procedures and concepts were developed, resulting in a reference concept in which the containment will essentially remain intact (in-situ concept). Over the top of the outer reactor vessel a disassembling area for remotely controlled tools will be erected that tightens on that vessel and can move down on the vessel according to the dismantling progress. (author)

MTR and PWR/PHWR in-pile loop safety in integration with the operation of multipurpose reactor - GAS

International Nuclear Information System (INIS)

Suharno; Aji, Bintoro; Sugiyanto; Rohman, Budi; Zarkasi, Amin S.; Giarno

1998-01-01

MTR and PWR/PHWR In-Pile Loop safety analysis in integration with the operation of Multipurpose Reactor - Gas has been carried out and completed. The assessment is emphasized on the function of the interface systems from the dependence of the operation and the evaluation to the possibility of leakage or failure of the in-pile part inside the reactor pool and reactor core. The analysis is refers to the logic function of the interface system and the possibility of leakage or failure of the in-pile part inside reactor pool and reactor core to consider the integrity of the core qualitatively. The results show that in normal and in transient conditions , the interface system meet the function requirement in safe integrated operation of in-pile loop and reactor. And the results of the possibility analysis of the leakage shows that the possibility based on mechanically assessment is very low and the impact to core integrity is nothing or can be eliminated. The possible position for leakage is on the flen on which one meter above the top level of the core, therefore no influence of leakage to the core

Measurement of the top quark mass

International Nuclear Information System (INIS)

Blusk, Steven R.

1998-01-01

The first evidence and subsequent discovery of the top quark was reported nearly 4 years ago. Since then, CDF and D0 have analyzed their full Run 1 data samples, and analysis techniques have been refined to make optimal use of the information. In this paper, we report on the most recent measurements of the top quark mass, performed by the CDF and D0 collaborations at the Fermilab Tevatron. The CDF collaboration has performed measurements of the top quark mass in three decay channels from which the top quark mass is measured to be 175.5 ± 6.9 GeV=c 2 . The D0 collaboration combines measurements from two decay channels to obtain a top quark mass of 172.1 ± 7.1 GeV/c 2 . Combining the measurements from the two experiments, assuming a 2 GeV GeV/c 2 correlated systematic uncertainty, the measurement of the top quark mass at the Tevatron is 173.9 ± 5.2 GeV/c 2 . This report presents the measurements of the top quark mass from each of the decay channels which contribute to this measurement

Top-emitting organic light-emitting diodes.

Science.gov (United States)

Hofmann, Simone; Thomschke, Michael; Lüssem, Björn; Leo, Karl

2011-11-07

We review top-emitting organic light-emitting diodes (OLEDs), which are beneficial for lighting and display applications, where non-transparent substrates are used. The optical effects of the microcavity structure as well as the loss mechanisms are discussed. Outcoupling techniques and the work on white top-emitting OLEDs are summarized. We discuss the power dissipation spectra for a monochrome and a white top-emitting OLED and give quantitative reports on the loss channels. Furthermore, the development of inverted top-emitting OLEDs is described.

Participative Design With Top Management

DEFF Research Database (Denmark)

Simonsen, Jesper

2004-01-01

meetings aimed at aligning top management with the supplier’s analysis. The article describes the MUST method’s anchoring principle and the technique of problem mapping supporting this principle. This participatory approach resulted in mutual learning processes with top management which is rarely reported...... on in the PD community. Top management participated by reviewing, challenging, and reformulating the IT designers’ central suppositions, assumptions, and hypotheses related to the causal relation between identified problems and suggested solutions....

Scoping calculations for design and analysis of large reactor vessels for liquid-metal fast breeder reactor (LMFBR) plants

International Nuclear Information System (INIS)

Fiala, C.; Kulak, R.F.; Ma, D.C.; Pan, Y.C.; Seidensticker, R.W.; Wang, C.Y.; Zeuch, W.R.

1982-01-01

Reactor vessels for commercial-sized LMFBR plants are quite large - ranging 40 to 70 ft in diameter and 50 to 70 ft in overall depth. These stainless steel vessels contain liquid sodium at relatively low pressures, but at high temperatures. The resulting thin-walled vessels present the structural designer and analyst with special problems, particularly in providing a balanced design to accommodate seismic loads, design basis accident loads, and thermal loadings. A comprehensive set of scoping calculations - though preliminary in detail and depth of design - provides substantial guidance to the vessel designer for subsequent design iterations. Emphasis is placed on the analysis of the large-diameter top closure of the vessel - the deck structure

Neutron physical investigations on the shutdown effect of small boronated absorbing spheres for pebble-bed high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Sgouridis, S.; Schurrer, F.; Muller, H.; Ninaus, W.; Oswald, K.; Neef, R.D.; Schaal, H.

1987-01-01

An emergency shutdown system for high-temperature gas-cooled pebble-bed reactors is proposed in addition to the common absorber rod shutdown system. This system is based on the strongly absorbing effect of small boronated graphite spheres (called KLAK), which trickle in case of emergency by gravity from the top reflector into the reactor core. The inner reflector of the Siemens-Argonaut reactor was substituted by an assembly of spherical Arbeitsgemeinschaft Versuchsreaktor fuel elements, and the shutdown effect was examined by installing well-defined KLAK nests inside this assembly. The purpose was to develop and prove a calculational procedure for determining criticality values for assemblies of large fuel spheres and small absorbing spheres

Use of LEU in the aqueous homogeneous medical isotope production reactor

Energy Technology Data Exchange (ETDEWEB)

Ball, R.M. [Babock & Wilcox, Lynchburg, VA (United States)

1997-08-01

The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.

Use of LEU in the aqueous homogeneous medical isotope production reactor

International Nuclear Information System (INIS)

Ball, R.M.

1997-01-01

The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Measurement and analysis of the neutron noise of the pool research reactor at IPEN

International Nuclear Information System (INIS)

Simoes, Graciete Pedro

1979-01-01

Variations in the neutron density or power of a nuclear reactor (the neutron noise) operating at nominally constant power are generally random and can only be described in terms of statistical parameters. Random variations in the power of a power reactor are produced by one or more driving functions. In this work the neutron noise of the pool reactor IEAR-1 (2 MW nominal power) has been studied using two compensated ionization chambers ( Westinghouse VJL6377) and related to three possible-driving functions, namely vibration of the control bar and reactor support bridge and the temperature of the water entering the core. The CIC detectors were located in rigid tubes in turn positively located in the reactor lattice plate. Conventional accelerometers were used. Temperature measurements were made with a NiCr/Ni thermocouple (wire diam ∼ 0.2mm) located 10 mm above the top of a fuel element. Although the correlation between the measured neutron signals was high ( > 0,4) for frequencies in the range 0 to 10 Hz no resonances were identified in the neutron noise. A significant correlation (> 0,4) between the control bar acceleration and the neutron flux was obtained in the frequency range 0 to 10 Hz. The measured correlation between the neutron noise and both the bridge vibration and the reactor water inlet temperature was insignificant. (author)

Top quark production at the LHC

CERN Document Server

Ferreira da Silva, Pedro

2016-01-01

Twenty years past its discovery, the top quark continues attracting great interest as experiments keep unveiling its properties. An overview of the latest measurements in the domain of top quark production, performed by the ATLAS and CMS experiments at the CERN LHC, is given. The latest measurements of top quark production rates via strong and electroweak processes are reported and compared to different perturbative QCD predictions. Fundamental properties, such as the mass or the couplings of the top quark, as well as re-interpretations seeking for beyond the standard model contributions in the top quark sector, are extracted from these measurements. In each case an attempt to highlight the first results and main prospects for the on-going Run 2 of the LHC is made.

Engineering solutions for a reflector change concept in the high-temperature reactor with pebble bed core and OTTO-fueling

International Nuclear Information System (INIS)

Kasper, K.J.

1975-06-01

In the field of reactor engineering an increasing tendency is visible towards a 'repairable reactor'. In the construction of the HTR with spherical fuel elements this fact should already be taken into account at an early stage. Additionally it is possible that in connection with the OTTO-fueling load conditions for the graphite reflector could result which are locally not far away from limiting values. Therefore the removability of the reflector is included in the reactor construction as an accompanying technical step of the physical lay-out of the core. The core arrangements, realized for HTR until recently, are discussed as well as the properties of the graphites used and the operating conditions in the reactors are stated. At the example of the PR 3,000 proposals are offered for the construction of a removable side and top reflector for a pebble bed reactor. Hereby a solution was found which, on one hand allows the changing of the reflector and on the other hand requires no significant increase of the costs for the reactor assembly. Moreover the requirements of reactor operation and of repairability are satisfied in an optimal manner. (orig.) [de

Argon-41 production and evolution at the Oregon State University TRIGA Reactor (OSTR)

International Nuclear Information System (INIS)

Anellis, L.G.; Johnson, A.G.; Higginbotham, J.F.

1988-01-01

In this study, argon-41 concentrations were measured at various locations within the reactor facility to assess the accuracy of models used to predict argon-41 evolution from the reactor tank, and to determine the relationship between argon gas evolution from the tank and subsequent argon-41 concentrations throughout the reactor room. In particular, argon-41 was measured directly above the reactor tank with the reactor tank lids closed, at other accessible locations on the reactor top with the tank lids both closed and open, and at several locations on the first floor of the reactor room. These measured concentrations were then compared to values calculated using a modified argon-41 production and evolution model for TRIGA reactor tanks and ventilation values applicable to the OSTR facility. The modified model was based in part on earlier TRIGA models for argon-41 production and release, but added features which improved the agreement between predicted and measured values. The approximate dose equivalent rate due to the presence of argon-41 in reactor room air was calculated for several different locations inside the OSTR facility. These dose rates were determined using the argon-41 concentration measured at each specific location, and were subsequently converted to a predicted quarterly dose equivalent for each location based on the reactor's operating history. The predicted quarterly dose equivalent values were then compared to quarterly doses measured by film badges deployed as dose-integrating area radiation monitors at the locations of interest. The results indicate that the modified production and evolution model is able to predict argon-41 concentrations to within a factor of ten when compared to the measured data. Quarterly dose equivalents calculated from the measured argon-41 concentrations and the reactor's operating history seemed consistent with results obtained from the integrating area radiation monitors. Given the argon-41 concentrations measured